Within the framework of studies on the fuel cycle and transuranic actinide management (Pu, Am, Cm), the CEA and ORANO launched an R&D program on fast molten salt reactors (MSR) in 2020. The aim was to assess their opportunities with respect to fuel cycle management and to offer insights into their technical feasibility. ARAMIS-P is an abbreviation of ‘Advanced Reactor for Actinides Management in Salt’ with P for plutonium; this project focused on the preliminary design of a small unit for plutonium conversion integrated into a spent fuel reprocessing facility. The goal was to avoid nuclear cycle issues like spent MOX fuel reprocessing, by investigating the possibility of providing a flexible plutonium conversion service (from high-grade plutonium to ex-MOX quality). The reactor power was fixed at 300 MWth, which is in the power range of advanced modular reactors (AMR). From the perspective of a reprocessing unit, the reactor footprint, the fuel salt hold-up, and the need to develop new chemical processes should be limited. A chloride-based fuel salt was selected due to its compliance with known spent fuel processes. A specific design with a high core power density and a 6-month batch feed-up strategy was chosen to limit the overall amount of fuel salt. The design process was also driven by the desire to make maximum use of proven technologies when available, as well as to implement a maintenance-by-design approach. This report presents the design methodology developed to produce a preliminary reactor sketch, to illustrate its burn-up performance, and finally to give insights into component design.

Chloride-salt fast-spectrum Molten Salt Reactors (MSR) are being increasingly studied globally, generally for potential fuel cycle advantages. Multiphysics modeling capabilities for MSR are also developed in order to analyze the dynamic behavior of such reactors during normal operation and safety transients. In France, the ARAMIS-P reactor was recently proposed, with a preliminary design carried out, while new transient modeling capabilities were also becoming available at CEA. This paper proposes to review existing results and introduces new ones on the application of multiphysics models to assess the transient behavior of ARAMIS-P. The applications were performed using different codes: R2P2 and HEROS for simplified studies directly in support of design, MOSAICS and CATHARE3 for system-scale studies of the plant, and TRUST-NK for a more precise analysis of the core dynamics. The models applied couple 0D, multi-0D, 1D and 3D scales for thermal-hydraulics, to point kinetics and 3D diffusion-based kinetics for neutronics. The performed studies include steady-state calculations at nominal and lower power, load-following normal operation, and accident transients: reactivity insertion, loss-of-fuel-flow with and without shutdown of fission power, and loss-of-fuel-flow in a station blackout (SBO) accident. These studies also facilitated code-to-code comparisons, revealing consistent trends in the predicted physics while identifying and analyzing discrepancies. The transient results show the stabilizing effect of the strong coupling in a MSR between neutronics and thermal-hydraulics for load-following and reactivity insertions, but also the difficulties raised with the adoption of tight temperature margins during the preliminary design. In addition, parametric studies have been performed to assess the impact of the fuel salt volume outside of the core on the reactor dynamics, and to assess the decay heat removal (DHR) conditions required for a passive management of the reactor once shutdown. Such multiphysics studies extending beyond the fuel circuit thus have potential to give feedback on the design process of ARAMIS-P.

Molten salt reactors (MSRs) are Generation IV nuclear reactor concepts gaining significant research interest worldwide. The present work proposes a methodology for the definition and optimization of the MSR operating conditions and control strategy. The proposed methodology employs a novel 0D steady-state model, the multiphysics system-scale MOSAICS (MOlten SAlt Incompressible Calculation System) model and a Global Sensitivity Analysis (GSA) method based on Hilbert-Schmidt Independence Criterion indices. The methodology was then applied to the ARAMIS (Advanced Reactor for Actinides Management in Salt) fast-spectrum chloride-salt burner reactor. Two alternative control strategies are proposed in order to achieve target margins to the salt freezing and structure material limit temperatures during normal operation, plus a 20 % power variation per minute load-following objective. The performance of the control strategies was assessed through comparison with the natural behavior during a load-increasing transient, as well as during Unprotected Transient OverPower (UTOP) and Station Blackout (SBO) accidents. Controlled load variation transients are observed to reduce overall temperature variation rates throughout the salt circuits, while the inclusion of a variable fuel flow from the reactor commands can further limit these fluctuations. However, the selected operating conditions exhibited insufficient margins to freezing or the materials limit temperature under unprotected accident conditions. GSA enabled the derivation of correlations to characterize the dynamic response of MSRs to the accidents. Based on the observed reactor response to the various transients, potential modifications to the ARAMIS design and control strategies are proposed.

The control rod withdrawal (CRW) transient is one of the most important safety issues for the early design phase of sodium fast reactors (SFR). CRW transients can cause core melting, which can lead to severe accidents. To ensure the integrity of the core, it is crucial to study CRW transients in detail since early design phase of the reactor. Two approaches are used in numerical modeling of transients at CEA. Reference codes (APOLLO3 (Schneider et al., 2016), GERMINAL (Lainet et al., 2019) for example) are used to perform detailed and precise calculations, recently these tools are being coupled to form multiphysics platforms such as described in Pascal et al. (2021); Li et al. (2023). Despite being robust and validated, this type of simulation is not adapted to design phases due to the need in computing resources it requires. The second type of approach, selected in this paper, is to use simplified, parametric tools with a limited validity range. We present here a new numerical tool called MORPHEE, which is capable of simulating CRW transients. MORPHEE is a multiphysics and multiscale tool that includes neutronics, thermal-hydraulics, and fuel behavior. The tool has been developed to provide a comprehensive understanding of the CRW transient and to assess the safety of SFR. It is based on MACARENa (Droin et al. 2020) and OCARINa (Herbreteau et al. 2018) tools. They have been validated for simulation of thermal-hydraulics and fuel pin behavior, respectively. MORPHEE uses finite-differences methods to solve thermal-hydraulics equations, neutron kinetics equations and mechanical equations. A machine learning model (Jeannin et al., 2022) is also used to predict fuel cladding heat transfer through the gap. The simulations have demonstrated the importance of considering the multiphysics and multiscale aspects of the CRW transient to predict the behavior of the reactor during an accident. In conclusion, MORPHEE is a numerical tool that can simulate the CRW transient in SFR. It can be used to study the safety of SFR during the early design phase and to optimize the design to ensure the safety of the reactor.

The objective of the paper is to study the potential behaviour of a power system with high share of nuclear and less thermal plants, in which variable RES insertion increases − for example the French case −, in order to determine the specifications for the design of a potential nuclear reactor with high “manoeuvrability”. Moreover, the flexible reactor may participate more in the supply − demand balance and in particular during large frequency fluctuations caused by the high variability of RES. The studies are carried out with the PowerFactory software, which make it possible to highlight specific needs regarding the power ramp for an “ideal” flexible nuclear reactor. Using a benchmark network, the Kundur “2 areas-4 machines”, the flexibility requirements are obtained as a function of the grid disturbances. For this purpose, the penetration of variable RES is progressively increased, while nuclear power is reduced and thermal power plants are totally suppressed. The study shows that a drop in RES production directly impacts the minimal frequency. A faster response speed of nuclear power makes it possible to restore this stability and return to normal operating conditions imposed by the grid operator. This paper describes therefore the process of obtaining the flexibility criterion for different cases of insertion of variable RES.

The massive penetration of renewable energy sources (RES) that are variable and not “dispatchable”, may weaken the power system supply-demand balance. Nuclear power plants (NPP) contribute in part to this daily and seasonal balance thanks to the “load-following” mode in France for example, but there are still limits to their use. These limits prevent a nuclear power modulation as efficient and quickly as the conventional thermal power plants. The need in terms of power ramps for nuclear in a constrained power system has been quantified in previous studies. Nuclear may compensate for the removal of thermal power plants, in order to fulfill energetic strategies of CO2 reduction. The possibility that nuclear reactors can achieve power ramps of significant values (>5%Pn/min) is put forward and could make possible to replace the services currently provided by thermal power plants. The objective of the study is then to use these power system requirements as the main input parameter for the modelling of a current simplified nuclear reactor capable of responding to frequency control within a specific hypothesis framework. In this paper, a French 1300 MW pressurized water reactor is modelled. Parametric studies are carried out in order to reveal technical and technological constraints when increasing electric power ramp. The study explores ways of design, which may influence reactor flexibility, such as the neutron parameter, Doppler coefficient, or the thermohydraulic parameter, delay in the primary loop.

Multiphysics modeling has become a central point in nuclear reactor simulation as it takes into account interactions between physical fields involved. Nuclear reactors are highly multi-physics systems as neutronics, thermal transfer, mechanics and hydraulics are interacting to produce and maintain the power. In this system, numerous issues come from material behavior. In the case of Control Rod Withdrawal (CRW) accident the main issues come with fuel behavior as the major risk to prevent is the fuel melting and spread in the core, leading to its partial or complete meltdown. In these situations, specialized codes are often used to predict properties and the state of the fuel by aggregating isolated models, each one corresponding to a single phenomenon. This type of approach is very efficient for understanding global sequence of the accident and predicting numerous physical variables of the problem. However it is often very expensive in calculation time. In this paper we focus on developing a fast tool based on machine learning models in order to speed up the calculation of specific variable of interest. We propose here a brief physical context of the study, a description of the method used to chose and to train models and finally an evaluation of models. Using this tool in a multiphysics scheme will add negligible penalty on calculation time. The model is focused on modeling fuel cladding heat exchange coefficient hgap based on data extracted from GERMINAL-V2 fuel performance code developed at CEA for SFR. Considering the strong non-linearity of the variation of the hgap, it appears that Random Forest models and AdaBoost present the smallest deviation, lower than 1% and the faster response, less than 10 mu s. However, the sensitivity of the hgap for highest burnups does not allow to get a metamodel with a deviation smaller than 20% for the application case.

Sodium cooled fast neutron reactors (SFR) are one of the selected reactor concepts in the framework of the Generation IV International Forum. In this concept, unprotected loss of cooling flow transients (ULOF), for which the non-triggering of backup systems is postulated, are regarded as potential initiators of core melting accidents. During an ULOF transient, spatial distributions of fuel, structure and sodium temperatures are affected by the core cooling flow decrease, which will modify the spatial and energy distribution of neutron in the core due to the spatial competition of neutron feedback effects. As no backup systems are triggered, sodium may reach its boiling temperature at some point, leading to local sodium density variations and making the transient fluctuate in a two-phase flow physics where thermal-hydraulics and neutronics may interact with each other. The transient phenomenology involves several physic disciplines at different time and spatial scales, such as core neutronics, coolant thermal-hydraulics and fuel thermo-mechanics. This paper presents the results of thermal-hydraulic/neutronic coupled simulations of an ULOF transient on the SFR project ASTRID. These coupled calculations are based on the supervisor platform SALOME to link the neutron code APOLLO3® to the system thermal-hydraulic code CATHARE3. The physical approach used by the coupling to describe the neutron kinetic is a quasi-static adiabatic one, updating the normalized spatial power distribution periodically by performing static neutron calculations, while a point kinetic model associated to a neutron feedback model calculates the power amplitude variations.

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume.

In order to improve passive safety of Sodium-cooled Fast Reactors The French Alternative Energies and Atomic Energy Commission (CEA) has proposed a new core design called CADOR - an SFR core with enhanced Doppler reactivity feedback. One of its most important design features is the introduction of solid moderating materials inside each fuel assemblies to slightly decrease the average neutron energy. The article focuses on development and validation of a neutronics calculation scheme able to produce accurate results in case of CADOR and other fast cores with moderating materials. The study uses two different fuel assembly models moderated by metallic beryllium and zirconium hydride (ZrH2) respectively The study includes discussion of neutron scattering treatment and different ways of spatial homogenization and energy condensations. The results indicate that the accurate scattering treatment leads to much better estimation of Doppler constant, especially in case of ZrH2 moderated core. By using combined deterministic-Monte Carlo calculation scheme we are able to quantify the biases on global reactivity, reactivity feedbacks and control rod worth. We demonstrate that spatial homogenization plays a more important role in case of moderated CADOR assemblies and thus preserving certain level of heterogeneity within fuel assemblies can lower the calculation bias significantly.

Generation-IV sodium fast reactors (SFR) will only become acceptable and accepted if they can safely prevent or accommodate reactivity insertion accidents that could lead to the release of large quantities of mechanical energy, in excess of the reactor containment's capacity. The CADOR approach based on reinforced Doppler reactivity feedback is shown to be an attractive means of effectively preventing such reactivity insertion accidents. The accrued Doppler feedback is achieved by combining two effects: (i) introducing a neutron moderator material in the core so as to soften the neutron spectrum; and (ii) lowering the fuel temperature in nominal conditions so as to increase the margin to fuel melting. This study shows that, by applying this CADOR approach to a Generation-IV oxide-fuelled SFR, the resulting core can be made inherently resistant to reactivity insertion accidents, while also having increased resistance to loss-of-coolant accidents. These preliminary results have to be confirmed and completed to meet multiple safety objectives. In particular, some margin gains have to be found to guarantee against the risk of sodium boiling during unprotected loss of supply power accidents. The main drawback of the CADOR concept is a drastically reduced core power density compared to conventional designs. This has a large impact on core size and other parameters.

This paper presents a comparison of homogenization techniques implemented in the APOLLO3 platform for transport core calculations: standard scalar flux weighting and new flux–moment homogenization, in different combinations with (or without) leakage models. Besides the historical B1-homogeneous model, a new B-heterogeneous one has indeed been implemented recently in the two/three-dimensional-transport solver using the method of characteristics. First analyses have been performed on a very simple Sodium Fast Reactor core with a regular hexagonal lattice. They show that using the heterogeneous leakage model in association with flux-moment homogenization strongly improves the prediction of keff and void reactivity effects. These good results are confirmed when the application is done to the fissile assemblies of the more complex CFV (Low Void Effect) core of the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project of sodium-cooled fast breeder reactor (Generation IV).

The ECRIX experiments aimed at demonstrating the transmutation capacity of sodium fast reactors reg a ding minor actinides (Americium, Neptunium, and Curium). These artificial atoms, produced in reactors, are mainly responsible for radiotoxicity and heat production of nuclear wa ste and thus concur to the complexity of storage fa cilities. Transmutation aims at reducing the amount of these products by ir radiating them with neutron beams in dedicated faci lities. The ECRIX-H experiment, performed from 2003 to 2006 in the Fren ch sodium cooled fast reactor Phénix, was focused o n Americium transmutation in a softened neutron spectrum. The e xperimental pin was filled with a mixed material co mposed of americium oxide and magnesia. It was placed inside a calcium hydride container whose goal was to soften the neut ron spectrum in order to enhance transmutation. Local neutron slowi ng-down due to hydrogen scattering reactions is a c omplex phenomenon which represents a challenge in terms of computer s imulation. Previous interpretation campaigns highli ted limitations of standard fast reactor calculation schemes regarding both flux and depletion calculations with signific ant deviations (up to 50%). This paper will discuss of a new interpretati on based on up-to-date calculation schemes using th e stochastic transport calculation code TRIPOLI4D® shipping a new depletio n calculation module.

Before the definitive shutdown of the prototype Phenix, a final set of experiments was performed to gather important data about the operation and safety of sodium-cooled fast reactors (SFRs).Among the accident sequences that are to be taken into account, inadvertent withdrawal of a control rod is considered. During operation at nominal power, such a sequence induces a general power increase and local deformations of the power shape. Afterward, local fuel temperature increases can thereby lead to fuel melting and clad failure.The quasi-static control rod withdrawal test was specially designed to gather local power data on fissile assemblies and to complete validation databases of neutronic codes. The maximal deformation of the power shape reached +/- 12% and was obtained when two control rods were shifted in opposite directions.The test analysis was conducted with the neutronics code ERANOS-2.2. Comparisons between calculated and measured values were satisfying. Most of the discrepancies in power estimation can be explained by measurement problems (heat transfer; sodium mixing).The association of ERANOS-2.2 and the nuclear library JEFF-3.1, presently used for the predesign phase of the ASTRID reactor, constitutes an acceptable predictive tool for local and integral parameter estimations in SFRs, specifically in the evaluation of the control rod withdrawal incident.

In 1989 and 1990, four very fast and high amplitude decreases of the measured neutron flux, known as 'AURN' events, occurred in the Phenix fast neutron reactor and were not well explained despite a significant investigation program carried out in the 90 s. The perspective of the Phenix end of life led the French Nuclear Energy Div. to reinitiate AURN studies in particular with dedicated Phenix tests. It comes out from simulations with improved fluid-structure interaction model of the core movement that an outward movement of the sub-assemblies can generate a power variation consistent with the AURN signals. 'AURN' Phenix tests confirm the abnormal thermal behavior of an experimental sub-assembly, a new A URN scenario has been proposed on this basis. (authors)

The French sodium cooled fast reactor Phenix was shut down in 2009 after 35 years of operation. Before decommissioning a final set of tests was performed. This paper focuses on the following core physics tests: - measurement of control rod reactivity worth by several methods (sub-critical, critical, rod-drop methods), - control rod shifting during a full power operation state, - measurement of individual subassembly reactivity worth (fresh and burned fuel and fertile, sodium hole), - simulation of a gas bubble crossing the core. The control rod measurement test has shown some discrepancies between the different measurement methods, mainly for the rod bank worth. Considering a macroscopic parameter (the reactivity loss estimation), NSMM method seems to produce the better results. The control rod shifting test has highlighted the impact of the spatial effects generated by control rods movements over the power map. These spatial effects, known as shadowing effects, can modify up to 15% the individual control rod worth. The tests concerning individual subassembly worth and gas bubble have permit to evaluate the impact of local perturbations on the reactivity. All these tests can be considered as successful and the ability of the European neutronic code for fast reactors, ERANOS 2.2,more » to reproduce complex and atypical configuration has been demonstrated by good agreement between measured and computed values. (authors)« less

The 250 MWe (140 MWe since 1993) PHENIX sodium cooled fast reactor was shut down on March 6th, 2009. Before decommissioning, a final set of tests were performed during the May 2009 – January 2010 period covering core physics, fuel behaviour and thermalhydraulics areas. Detailed analysis of the tests results is ongoing. It will be used for the extension of the validation of ERANOS and DARWIN codes for core physics, TRIO_U and CATHARE for thermalhydraulics and GERMINAL for fuel behaviour.In addition, the program included two tests related to the comprehension of the four negative reactivity transients (AURN in French acronym) experienced during the reactor operation in ‘89 and ‘90 and not yet fully explained.This was also a great opportunity to involve young engineers in the different processes like the design of the tests, their carrying out, and the analysis of the results.The standard instrumentation of the reactor was completed by specifically designed devices.

We explore experimentally and numerically the physics underlying the optimization of high-order harmonic generation by intense laser pulses, whose temporal profile is tailored by a learning genetic algorithm. Based on a large set of optimization data obtained under different generation parameters, we show that the algorithm converges toward a class of very special profiles on the leading edge of the pulse. The behavior of the harmonic signal is then compared with theoretical simulations based on the time-dependent Schrodinger equation, allowing one to identify separately the role of microscopic and macroscopic phenomena in the temporal dynamics of high-harmonic generation and optimization.

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)