Development of an uncertainty propagation method of function-typed input data applied to the decay heat calculation

Characterising the energy released by the disintegration of the radionuclides present in spent nuclear fuel is essential for the design, safety and analysis of storage, transport and disposal systems. Few measurements of this decay heat are available today. In addition, the available experimental values do not cover the wide spectrum of possible combinations between parameters such as discharge burn-up rate, 235U enrichment, cooling time, fuel design parameters, or operating conditions. The estimation of decay heat is therefore mainly based on calculation codes.
The evaluation of the uncertainty associated with the estimation of decay heat is important to achieve reliable predictions. Many efforts have been made to properly evaluate biases and uncertainties coming from nuclear data such as cross sections. The number of studies concerning uncertainties of an epistemic nature (uncertainty in the manufacture of some components, error in reading or adjusting mobile structures, etc…) is comparatively small. Among the latter, while the treatment of complex dependencies of scalar input parameters is well taken into account today, functional-type dependencies, i.e., expressed in the form of a function, are very little explored.
While uncertainties arising from the processing of fixed input parameters, such as fuel manufacturing parameters, independent of time, are quite well covered, the uncertainties coming from the processing of variable (or functional) parameters, such as operating history, evolving during reactor operations, are not. Irradiation history actually brings together several inter-correlated quantities (operating power, absorber movements, core evolution …), subject to modifications over time and influencing the value of numerous observables of interest, including decay heat. The models used today in industrial simulation tools do not make it possible to estimate this impact and to infer a validated uncertainty.

This research work will investigate the impact on decay heat of the uncertainties associated with input parameters having functional dependencies. We will particularly focus on the irradiation history of the reactors (PWR type). A first part of the work will be dedicated to the development of a substitution model for decay heat estimation and quantification of uncertainties of a functional nature. The second part will be devoted to the development of a sensitivity analysis method. Finally, a third part will concern the development of an inverse method for quantifying the uncertainties coming from irradiation modelling.
The doctoral student will be hosted in a reactor physics research unit of the CEA IRESNE located in Cadarache where he will collaborate with other doctoral students and specialists in the field.

Development of advanced optimisation methods for nuclear power scenarios

The study of possible nuclear fleet evolution is done through scenario calculations. A scenario models precisely all material flows within the fuel cycle, starting with raw material extraction, following with fuel fabrication, fuel irradiation inside the reactor, spent fuel cooling, fuel reprocessing and waste disposal. The scenario is thus a great tool for decision making. However, a scenario is really dependant on the set of hypotheses considered, that are affected by deep uncertainties. The current way to perform scenario calculation is not well suited to manage such hypotheses changes due to uncertainties.

A new field of research has emerged to deal with these deep uncertainties : the study of scenario robustness and resilience. The objective is no longer to quantify the performances of a precise scenario, but its ability to be modified to answer to the objective or constraint change (such as an installed power variation). To do so, it is necessary to launch several thousands of calculations, among which a large part are not viable.

The goal of this thesis work is to investigate the optimization methods used in logistics in order to build efficient methods to quickly build scenario inputs. The generated inputs should lead to optimal scenarios for a set of given objectives. Then, it would be possible to identify the scenarios that are able to answer to several objectives and assess whether they can be adjusted to answer to new constraints. In other words, this thesis is another step towards the production of resilient scenarios against future uncertainties.

Batch fuel management of molten salt nuclear reactors and consequences on the fuel cycle

Many molten salt reactor concepts rely on so-called continuous salt management, which consists of continually inserting/removing fuel salt into/from the core to compensate for the loss of reactivity due to fuel depletion. In this PhD thesis, we propose to reconsider the molten salt reactor conceptual design in a so-called batch fuel management strategy that involves taking and loading a fraction of the volume of the core after a certain irradiation time and during a reactor maintenance. Contrary to a continuous strategy, the aim of this batch refueling strategy is to take into account technological constraints external to the reactor. The implications of a batch-fueled molten salt reactor with regard to core performances, cycle constraints and salt chemistry constitutes a largely unexplored area of research.

The doctoral student will first focus on evaluating the neutronic impacts of batch management with sensitivity studies (minor actinides burning performances, fuel regeneration performances, cycle time, volume/mass supply). The doctoral student will then focus on the cycle aspects of a molten salt reactor integrated into a given nuclear fleet, including the fresh fuel salt fabrication (actinides solubility) and the used salt reprocessing (process, cooling time, process time) with nuclear scenario calculations. The approach will evaluate the relevance of a batch-fueled molten salt reactor and will be applied to different reactors (burner, breeder) in comparison to continuous salt management.

The thesis will allow the candidate to develop skills in the conceptual design of a fourth generation reactor. He/she will be part of the scientific community working on such complex systems, which opens the door to a job in a R&D lab.

Design and realization of a high-temperature optical neutron detector. Application to an experimental program in the JOYO reactor

As part of the development of fourth-generation sodium-cooled fast reactors, the CEA/IRESNE Dosimetry and Instrumentation Laboratory is working on innovative neutron measurement systems capable of operating at temperatures of the order of 600°C, and insensitive to the parasitic phenomena that occur under these conditions. Recently, a new type of optical signal neutron detector (ODN) has been developed at the laboratory. Despite a more complex signal interpretation, this instrument has several advantage: it can be miniaturized and it is intrinsically insensitive to problems of partial discharges and leakage currents that occur in ionization chambers at high temperature.
We propose to pursue the theoretical and experimental development of ODNs to adapt them to high temperatures. The PhD student will further develop the modelling tools already available in the laboratory for simulating the detector response. The work will investigate heavy ion-noble gas interaction cross sections, also a radiative collisional model to predict emission spectra and their temporal dynamics. Part of the work will involve dimensioning a high-temperature prototype and testing it in the JSI TRIGA reactor. Ultimately, the detector will be qualified in the JOYO research reactor as part of a broader experimental program.

Analysis and multi-scale thermal-hydraulic simulation of design transients of an innovating nuclear-to-heat reactor concept

The System optimization and pre-design Laboratory of CEA/IRESNE at Cadarache works on innovating nuclear reactor concepts in order to decarbonize all industry and urban sectors (flexible electricity, heat, cool, synthetic fuel, hydrogen). One of those innovating concept is the ARCHEOS passive water reactor dedicated to heat supply and designed to be intrinsically safe and simple to operate. The main challenge of this research is to understand and analyse the thermal-hydraulic behaviour of this reactor that fully operates in natural circulation, which is clearly an innovation in the domain. The PhD student will first identify normal and accidental scenarios and simulate them at the reactor scale. Thoughts for design improvements could emerge as a result of this research. Those simulations will be associated to a deep physical analysis of thermal-hydraulic phenomena that can play a role during the studied scenarios. An appropriate modeling (from 1D to porous 3D to 3D CFD) is to be found to capture the thermal-hydraulic phenomena of importance. This will be performed using CATHARE3 and Neptune_CFD tools. Working on such an innovating nuclear reactor concept represents a great opportunity for a PhD student. This experience will be provide the student with a solid background on various topics such as: nuclear safety, innovating reactor design, multi-scale thermal-hydraulic simulation, reactor physics in transient regimes as well as a solid knowledge on the CATHARE3 code which is widely used in French nuclear industry and research and the reference system code for many projects in the nuclear industry.

Thermal-hydraulic simulations of turbulent flows using the immersed boundary method for innovative nuclear reactor safety devices

The Cadarache CEA/IRESNE R&D unit on nuclear system modelling investigates passive safety devices to minimize the flow of water out of the reactor vessel of a pressurized water reactor and to manage the water reserves available for safety injections, in the event of a loss-of-coolant accident. These devices, such as in-vessel flow limiters or advanced accumulators, operate on the principle of hydraulic diodes, to prevent or delay core dewatering and its possible degradation.
The subject of this thesis is the Penalized Direct Forcing numerical modeling of turbulent thermal-hydraulic flows under various spatial discretizations. The technique of introducing a direct forcing term into the Navier-Stokes equations allows obstacles to be taken into account in an incompressible flow. It combines projection and velocity penalization methods. It leads to a natural treatment of boundary conditions for pressure correction at the edges of obstacles. The doctoral student will build on the achievements of two recent PhD theses, dedicated to the simulation of laminar or turbulent flows using the PDF method in Finite-element spatial discretization using a scalar model of turbulence. He/she will extend this work to other spatial discretizations (Finite-Element Volumes, Finite-Difference Volumes) and to turbulence-model evolution equations.
The computational methodology will be verified/validated on test cases and applied to the simulation of hydraulic diodes envisaged for passive safety systems in nuclear reactors.
The doctoral student will develop competence in thermohydraulics simulation and numerical methods.

Monte Carlo simulation of the reactor transfer function to improve neutron noise measurement analyses

The neutron population in a reactor fluctuates due to the random nature of neutron emission and various sources of mechanical vibrations, which can impact macroscopic neutron cross sections. The reactor can be seen as a system with a transfer function that connects an excitation (such as a vibration or the random nature of neutron emissions from fission) to the neutron population. The study and measurement of this transfer function allow us to deduce essential neutron parameters related to the kinetics of delayed neutron emission or even the source of thge vibrations. However, the theoretical expression of this transfer function is often based on the kinetics of the point reactor, which in some cases does not reliably exploit the measurements.
In this thesis work, we propose to study various extensions of the neutron transfer function formalism using Monte Carlo simulations. First, we will simulate fluctuations using a simplified C++ model to confirm the assumptions of theoretical equations for "neutron noise" that can be used to "measure" the effective fraction of delayed neutrons. We will then seek to optimize the positioning of detectors in a reactor and interpret certain effects related to positioning already observed in past experiments conducted by CEA.

Innovative modeling for multiphysics simulations with uncertainty estimates applied to sodium-cooled fast reactors

Multiphysics modeling is a powerful tool for analyzing nuclear reactors, but the uncertainty propagation between disciplines is often disregarded. This PhD thesis proposes innovative approaches to improve the accuracy of multiphysics modeling by accounting for these uncertainties. The primary goal is to propose optimal modeling approaches tailored to diverse accuracy requirements. This information is of prime interest to researchers and industry professionals involved in the development and utilization of multiphysics models. Specifically, the thesis will assess various uncertainty propagation techniques applicable to multiphysics simulations. This involves exploring surrogate modeling through avenues like reduced-order modeling and polynomial chaos expansion. The goal is to identify and categorize input parameters with the most significant impact on system outputs, irrespective of their physical domain. Subsequently, uncertainty propagation will be executed using two core modeling types: a ‘high-fidelity’ model based on the CEA's reference simulation tools and a ‘best-estimate’ model accounting for the "industrial" objective of the calculations). The similarities and differences between these approaches will be analyzed to assess model biases. These uncertainty evaluations employing the above methods will be tested on an extensive set of experiments performed in SEFOR, a sodium-cooled fast reactor, representing a diverse range of experimental data for various reactor conditions.

Measurements of correlations between fission observables to constrain fission fragment de-excitation models

The Laboratory of physics studies at CEA/IRESNE Cadarache has been developing for about ten years a Monte-Carlo code called FIFRELIN capable of simulating the nuclei (fragments) resulting from a fission reaction and predicting their de-excitation by neutron, gamma and electron emission. Thanks to this tool, the calculation of a large number of "fission observables" needed in reactor physics becomes possible. However, it is necessary to validate the models implemented in the code. One way of doing this is to carry out so-called multi-observable experiments that FIFRELIN will try to reproduce by refining the models and/or the parameters of these models. These multi-observable experiments consist in measuring in coincidence the two fission fragments as well as the neutrons and prompt gamma emitted by these fragments, in order to determine the correlations between these observables. This is precisely the purpose of the VESPA++ device, recently developed by the Joint Research Center of the European Commission located in Geel, Belgium (JRC-Geel).
Two 18-month phases are planned during this thesis. A first experimental phase during which the student will be based at JRC-Geel. He/She will get familiar with the functioning of the VESPA++ device, will participate in the experiments and analyze the raw data collected. A second phase will be dedicated to the improvement of the models implemented in FIFRELIN in order to reproduce the experimental results. During this second phase, the student will be based at Cadarache.

Prompt fission neutron multiplicity in the Resolved Resonance Region of 239Pu

Prompt fission neutron multiplicity in the Resolved Resonance Region of 239Pu

This thesis investigates the role of the (n,gf) process on the number of prompt neutrons emitted per fission (multiplicity “nubar”) in the resolved resonance region of Pu-239 (from 0.3 to 100 eV). The emission of gamma rays prior to fission may explain the observed reduction in nubar in this energy region. In a first step, time-of-flight experiments are to be carried out at JRC's GELINA facility in Geel, Belgium. A second step of the thesis will be to use the FIFRELIN code developed in our Cadarache laboratory to simulate fission events. For resonances with a strong (n,gf) component, the gamma emitted before fission should significantly modify the spectrum of prompt fission gammas emitted by the fragments, due to a decrease in the excitation energy in the system. This phenomenon of gamma emission before fission is not yet taken into account in the FIFRELIN code, the student will have to implement a model to do so. A simple solution would be to model the process in two stages: emission of a gamma by the compound nucleus (model already implemented) and then fission of the residual nucleus (model also implemented). As part of this research work, the doctoral student will develop competence in theoretical and experimental nuclear physics