French Atomic Energy Commission Research Articles

LES investigations of turbulent heat transfer structures with exponential power escalation

The BORAX-type accident is a key scenario in the safety design of pool-type experimental nuclear reactors. It considers a reactivity insertion large enough to initiate an exponential power escalation with a period as small as tens of milliseconds, necessitating effective heat removal by the coolant system. This study presents a computational analysis of single-phase transient heat transfer under thermal-hydraulic conditions relevant to such reactors, focusing on turbulent channel flow between planar fuel plates with highly subcooled water. Our investigation uses Large Eddy Simulation (LES) performed with TrioCFD, the open-source CFD code developed by the French Atomic Energy Commission (CEA). Several modeling options are tested: incompressible flow with temperature as a passive scalar, incompressible flow with varying viscosity, and quasicompressible flow with temperature-dependent viscosity and mass density. To evaluate the accuracy of our physical models, all simulations were performed under uniform conditions, using the same experimental parameters, mesh, and numerical strategies. The specific examined scenario involves a highly subcooled turbulent channel flow at moderate pressure, experiencing an exponential power increase. We ensure detailed turbulence resolution at the Batchelor scale in the wall normal direction close to the heating wall. This communication presents a comparative study of LES simulations of the single-phase flow under exponential power excursion at the wall. Subsequent analyses focused on the impact of different velocity-temperature couplings on the prediction of wall heat transfer, incorporating comparisons with experimental infrared thermometry measurements. Early in the transient phase, temperature peaks are confined within turbulent streaks, as reported in the existing literature. However, as the exponential power transient unfolds, the viscous layer and streak structures destabilize, leading to a more dispersed distribution of hot spots. Preliminary findings suggest a transition in heat transfer modes from turbulence-driven to conduction during the final stages of the single-phase transient.

Multi-physics code coupling methodology for the simulation of accidental transients in sodium cooled fast reactors

The French Alternative Energies and Atomic Energy Commission (CEA) has played a substantial role in advancing fourth-generation reactor technologies, including Sodium-cooled Fast Reactors (SFR), which are anticipated to offer enhanced efficiency, adaptability, and sustainability compared to existing nuclear technologies. In the context of safety assessment, it is crucial to precisely simulate accidental scenarios like Unprotected Loss Of Flow (ULOF) transients. These transients entail intricate multi-physics interactions within the reactor, particularly within the core, encompassing thermal-hydraulics, reactor physics, and fuel performance (e.g. changes in densities or temperatures will lead to neutronics effects, thus power, then causing feedbacks onto said densities and temperatures).In order to capture those complex phenomena, a multi-physics coupling approach for SFRs, which combines the CATHARE3 code for system scale thermal-hydraulics, the APOLLO3 code for reactor physics, and the GERMINAL V2 code for nuclear fuel performance has been developed at CEA in the recent years.This paper presents a comprehensive methodology of such coupling and its integration into a Verification, Validation and Uncertainty Quantification (VVUQ) process. The first part of the paper provides an introduction to the simulation tools used for the coupling. The following section explains the multi-physics coupling approach, facilitated by CEA’s Collaborative Code Coupling Platform (C3PO), which enables code integration through a Python interface to create a comprehensive simulation model. A detailed examination of the coupling algorithm and data transfer mechanisms is also provided.The generality and reliability of this approach are demonstrated by simulating two accidental transients: the Unprotected Loss Of Flow (ULOF) in the ASTRID reactor and the Loss Of Flow WithOut Scram (LOFWOS) in the Fast Flux Test Facility (FFTF), both corresponding to a primary pump trip without control rods falling. The results show that the proposed multi-physics coupling scheme is effective and robust for SFR simulations. The coupled model enables a more accurate and realistic representation of the complex phenomena occurring in the reactor during an accident.This research underscores the significance of multi-physics coupling in the analysis of SFRs and serves as a valuable point of reference for future SFR studies. The proposed approach offers an efficient way to investigate SFR responses during accidents or incidents, thereby advancing the development of dependable and precise simulation tools for SFR design and evaluation.

Long-Term Infrasound Sensor Calibration and Characterization

Abstract Previous testing has shown that infrasound sensors deployed in the field can exhibit notable deviations from their nominal, lab-based calibrations. These variations may be due to changes in environmental conditions, long-term sensor drift, or other unresolved features. In early 2018, we installed two identical infrasound elements with five infrasound sensors at each element (Chaparral M50A, Chaparral M64LN, CEA/Martec MB2005, CEA/Seismowave MB3a, and Hyperion IFS-5113A). These sensors were accepted or under consideration for use in the International Monitoring System network of the Comprehensive Nuclear-Test-Ban Treaty. Each element had all sensors connected to a single digitizer and port to the atmosphere. We also recorded instrument enclosure air temperature and humidity and external air temperature. Using the MB2005 as the reference, we examine the relative sensor response (both magnitude and phase) as a function of time and frequency and compare it with quarterly laboratory calibrations and environmental conditions. We find that the magnitude response for all sensors exhibits some variability in both the lab and field, with the amplitude variations often >5%. The field-based variations are more severe and occur on both long-term (months) and short-term (hours) timescales. Short-term variability correlates with changes in environmental conditions and is considerable (up to 25%) for the Chaparral M50A and noticeable (∼5%) for the French Alternative Energies and Atomic Energy Commission (CEA) MB3a. Long-term magnitude variability for the Chaparral M50A was up to 50% and does not clearly correlate with environmental conditions. The other sensors show some long-term magnitude offsets, but they have relatively stable responses in the conditions we examined. The MB3a also displays some frequency-dependent magnitude variability and shows a minor dependence on temperature. Phase estimates are stable and near zero for all sensors tested. These results strongly suggest sensor response and variability due to environmental conditions should be considered in future infrasound data interpretation and sensor selection and development.

Overview of multiphysics R&D activities at the CEA/IRESNE institute

The major mission of the IRESNE (Research Institute on Nuclear Systems for low carbon Energy production) institute of the DES (Energy Division) at CEA (French Alternative Energies and Atomic Energy Commission) is to study current and future nuclear reactors integrated into low-carbon energy systems. To fulfill this mission, the IRESNE’s main activities rely on the development and use of advanced physical models, multiphysics tools including efficient coupling schemes, and dedicated experimentation. Approaches to nuclear systems using multiphysics simulations are now required, in order to better understand the complex intrinsic features of the coupled phenomena.Hence, IRESNE addresses eight different multiphysics research fields: fuel fabrication, fuel behavior under irradiation, fuel assembly integrity under fluid–structure interaction, contamination of fluid circuits, sodium technology, behavior of reactors, corium and fission products, and radionuclide transfer to environment. A general overview of IRESNE’s current multiphysics activities in these eight fields is provided in this paper. It focuses on experimentation, modeling, simulation and tool qualification, corresponding to the Verification Validation and Uncertainty Quantification process. Moreover, IRESNE’s key challenges to be met in the next 5 to 10 years, to improve and validate multiphysics Modeling and Simulation, are also discussed in this contribution. The ultimate goal is to get closer to the development of a digital reactor, core and fuel, still progressing on applied mathematics, high-performance computing, and understanding and modeling of physical mechanisms.

A New Calculation Strategy for Molten Salt Reactor Neutronic–Thermal-Hydraulic Analysis Implemented with APOLLO3® and TRUST/TrioCFD

Molten salt nuclear reactors (MSRs) constitute a promising technology to produce safe, reliable, abundant low-carbon energy. To design MSR systems and perform safety analyses on them, numerical simulation is a powerful tool. Here, we implemented a coupling between several solvers of the deterministic neutronics code APOLLO3® (the MINARET SN transport and the MINOS diffusion and SPn-simplified transport solvers) and the computational fluid dynamics (CFD) code TRUST/TrioCFD, both developed at the French Alternative Energies and Atomic Energy Commission (CEA). The code coupling is orchestrated using the dedicated C3PO library of the open-source SALOME platform. A new code-coupling strategy is employed whereby the delayed neutron precursor concentrations are computed by the CFD code, which eases the use of traditional deterministic neutronics codes. We verified the correctness of our implementation by performing a numerical benchmark dedicated to fast spectrum MSRs originally devised by the French National Center for Scientific Research. The numerical results we obtained are in excellent agreement with those obtained by recent MSR-dedicated multiphysics simulation tools. This study provides a new convenient neutronic–thermal-hydraulic coupling strategy for MSR core simulation.

THE ROLE OF SOVIET JEWS IN BUILDING THE ISRAELI NUCLEAR ARSENAL (THE DIMONA REACTOR AS A MODEL) NINEVEH EDUCATION DIRECTORATE

The Israeli nuclear file is considered one of the biggest secrets of nuclear life in the world, if (Israel) refuses, since its establishment,to be included in the periodic inspections conducted by the Atomic Energy Agency on nuclear facilities in the world, just as its government refuses to sign the Nuclear Non-Proliferation Treaty signed by the Arab countries. Which makes the Israeli nuclear file (out of coverage). (Israel) possesses a basic nuclear reactor، as it was the first step it took regarding the establishment of a nuclear force in late (1948)AD and early (1949)AD. The most important and most dangerous effective step taken by (Israel) on the road to producing nuclear weapons for nuclear energy in(Israel)was its establishment of(the Dimona reactor.). This reactor was built according to French designs drawn up by the French atomic Energy Commission as part of the French-Israeli efforts to coordinate in the field of atomic cooperation. The Dimona reactor is located on the desert road between Beersheba and Sodom، in the Negev Center for Nuclear Research. It consists of ten buildings. It contains a nuclear reactor and a reprocessing plant. Plutonium This place was chosen because of its proximity to the raw materials (uranium) found in the Negev desert. The costs of building the reactor were estimated at about (130) million dollars, and its capacity was (150) megawatts, and it was able to produce (8) kilos of plutonium annually. The reactor operates with enriched uranium. (90%), which is a reactor that produces plutonium، which is the main element for the production of the fissile atomic bomb, since the Jews of Russian origins had a great and most prominent influence in the establishment of the Israeli nuclear program, as they pushed the Israeli nuclear file into existence. And they put (Israel) at the forefront of the nuclear countries

Development of a 3D APOLLO3® Neutron Deterministic Calculation Scheme for the CABRI Experimental Reactor

CABRI is an experimental reactor at the French Alternative Energies and Atomic Energy Commission (CEA) used to study fuel behavior during reactivity insertion transients. As these transients have a high level of multiphysics, it is important to develop suitable modeling and simulation tools to simulate them in order to be able to optimize testing and control of experimental conditions. This paper focuses on the development and validation of the neutron deterministic APOLLO3® calculation scheme that is included in the CABRI neutronic/thermal-hydraulic multiphysics coupled simulation tool; it represents the first stage of a stepwise validation process for the CABRI multiphysics simulation tool. The neutron calculation scheme is based on a classical two-step approach. The first step consists of a 281-energy-group flux calculation with the TDT-MOC (Method of Characteristics) solver for nuclear data space and energy collapsing for the different CABRI assembly clusters. The biases on a two-dimensional (2D) core neutron calculation due to self-shielding correction and collapsing on a restricted pattern are investigated by means of comparison with a direct full 2D calculation on a quarter core. The second step relies on a three-dimensional (3D) core calculation. Two approaches are presented. The first one consists of a best-effort approach corresponding to a 3D pin-scale description of the core, performing a transport calculation with the SN solver MINARET. And, the second one, a best-estimate approach, which will be implemented for kinetics calculations, relies on solving a simplified transport SPN equation in the solver MINOS with an exact 3D cell description of the core. The best-estimate calculation scheme is then used to analyze three experimental CABRI transients. A stepwise validation process is followed to quantify the calculation biases on physical parameters such as reactivity, reaction rates, and total core power at each step using static reference calculations with the stochastic code TRIPOLI4® or transient experimental data. The next development stage toward a multiphysics calculation scheme will be implementation and validation of coupling with a core thermal-hydraulic model.

Safety-oriented SFR core design: Methodology and application to transient bifurcations and stability maps analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commission), a methodology is proposed to early consider safety requirement in the undergoing reactor design process. Before the use of mechanistic tools (CATHARE, SIMMER, EUROPLEXUS, etc.) whose input deck elaboration requires an advanced knowledge of the reactor design, the methodology proposed in this article involves several physical tools simulating phenomena likely to govern the choice of design parameters. These tools are mostly based on reduced order models (ROM), meaning that they mainly involve low-dimensional modelings (mostly 0D and 1D) that are validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local flow blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, this fast-running platform makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientations for future R&D studies. In this article, a focus is firstly made on some scenario bifurcations to illustrate the platform capabilities. The boiling onset and possible reactor state stabilization are studied. The possibility of primary power excursion depending on the core design and the fuel vaporization possibilities are assessed. This platform is also of great help in order to rapidly compare several core designs when facing accidental transients. An example is given by comparing a CADOR core design to an ASTRID-like core design through flow stability maps in natural circulation conditions obtained by the fast-running tool platform. These applications demonstrate the efficiency of the presented methodology integrating safety at the very first design stages, and at facilitating the safety-oriented design of SFRs.

Microbial recycling of lithium-ion batteries: Challenges and outlook

Joseph Jegan Roy started his research career at the National Institute of Interdisciplinary Science and Technology (CSIR–NIIST) in Trivandrum, India, and received his PhD from the University of Kerala, India. Currently, he is a senior researcher in the SCARCE program at the Energy Research Institute @ NTU ([email protected]), Nanyang Technological University (NTU), Singapore. His current research centers on recycling spent lithium-ion batteries through biological, direct recycling techniques and waste-for-waste strategies. In addition, he is exploring the possibility of reutilizing recovered materials in novel applications of new electrochemical energy storage. Norazean Zaiden is a graduate of the Singapore Centre for Environmental Life Sciences Engineering (SCELSE, NTU, Singapore). Her interest in applied microbiology leads to biofilm matrix development for bioremediation and resource recovery applications through synthetic biology. In addition, she continues her journey in SCELSE as a research fellow under an industry collaboration studying anti-biofouling agents' effects on biofilms. Minh Phuong Do is a research fellow (materials science) with years-long experience in lithium/sodium ion batteries at Energy Research Institute at NTU, Singapore. She received her PhD from NTU in 2020, specializing in electrochemistry. Phuong uses this expertise in developing and optimizing a safe and stable battery system. Her current research focus is on the green closed-loop hydrometallurgical process for recycling spent lithium-ion batteries using biomass and regeneration of high-quality battery materials. Bin Cao is an associate professor in the School of Civil and Environmental Engineering and a principal investigator at the Singapore Centre for Environmental Life Sciences Engineering, NTU, Singapore. He received his PhD from the National University of Singapore and later worked as a postdoctoral research associate at Washington State University and Pacific Northwest National Laboratory. His research interest centers on biofilm-mediated microbial processes, with the goals of understanding environmental biofilm processes and applying the knowledge and insights to harness the power of beneficial biofilms and combat detrimental biofilms. Madhavi Srinivasan is a professor and executive director of the Energy Research Institute @ NTU ([email protected]) and NTU Sustainability office. Her research focuses on synthesis, fabrication, and application of nanoscale materials/architectures in improving the performance of electrochemical energy storage devices such as advanced lithium-ion batteries, supercapacitors, sodium ion batteries, multivalent aqueous Zn/Al batteries and recycling of e-waste. She is the co-director of SCARCE (Singapore-CEA alliance for research in circular economy), a joint lab, in collaboration with French Alternative energies and atomic energy commission (CEA, France), focusing on recycling of e-waste. Joseph Jegan Roy started his research career at the National Institute of Interdisciplinary Science and Technology (CSIR–NIIST) in Trivandrum, India, and received his PhD from the University of Kerala, India. Currently, he is a senior researcher in the SCARCE program at the Energy Research Institute @ NTU ([email protected]), Nanyang Technological University (NTU), Singapore. His current research centers on recycling spent lithium-ion batteries through biological, direct recycling techniques and waste-for-waste strategies. In addition, he is exploring the possibility of reutilizing recovered materials in novel applications of new electrochemical energy storage. Norazean Zaiden is a graduate of the Singapore Centre for Environmental Life Sciences Engineering (SCELSE, NTU, Singapore). Her interest in applied microbiology leads to biofilm matrix development for bioremediation and resource recovery applications through synthetic biology. In addition, she continues her journey in SCELSE as a research fellow under an industry collaboration studying anti-biofouling agents' effects on biofilms. Minh Phuong Do is a research fellow (materials science) with years-long experience in lithium/sodium ion batteries at Energy Research Institute at NTU, Singapore. She received her PhD from NTU in 2020, specializing in electrochemistry. Phuong uses this expertise in developing and optimizing a safe and stable battery system. Her current research focus is on the green closed-loop hydrometallurgical process for recycling spent lithium-ion batteries using biomass and regeneration of high-quality battery materials. Bin Cao is an associate professor in the School of Civil and Environmental Engineering and a principal investigator at the Singapore Centre for Environmental Life Sciences Engineering, NTU, Singapore. He received his PhD from the National University of Singapore and later worked as a postdoctoral research associate at Washington State University and Pacific Northwest National Laboratory. His research interest centers on biofilm-mediated microbial processes, with the goals of understanding environmental biofilm processes and applying the knowledge and insights to harness the power of beneficial biofilms and combat detrimental biofilms. Madhavi Srinivasan is a professor and executive director of the Energy Research Institute @ NTU ([email protected]) and NTU Sustainability office. Her research focuses on synthesis, fabrication, and application of nanoscale materials/architectures in improving the performance of electrochemical energy storage devices such as advanced lithium-ion batteries, supercapacitors, sodium ion batteries, multivalent aqueous Zn/Al batteries and recycling of e-waste. She is the co-director of SCARCE (Singapore-CEA alliance for research in circular economy), a joint lab, in collaboration with French Alternative energies and atomic energy commission (CEA, France), focusing on recycling of e-waste.

CHICADE nuclear facility: a collaborative technological platform, dedicated to the expertise and characterisation of nuclear wastes

CHICADE (CHimie CAractérisation DEchets – Chemistry CAracterization Wastes) is one of the nuclear facilities of the Energy Division of the French Alternative Energies and Atomic Energy Commission (CEA/DES). The CEA/DES is responsible for structuring and piloting the research programmes on energy at CEA. It involves its own institutes of research and those of other divisions. CHICADE is part of the Directorate for Nuclear Dismantling, Services and Waste Management. The Laboratory of Expertise and Destructive characterization is set up in the Basic Nuclear Facility – N° 156 called “CHICADE” where heavy equipment is used. The laboratory brings together both skills and means of characterization, using destructive methods on nuclear waste packages. It also carries out measurements on the whole waste package (gas release measurements, leaching tests). After a short presentation of the CEA and the Cadarache Centre, this publication aims to present the CHICADE facility and present the types of expertise on nuclear waste that are conducted there, i.e., measurement of the diffusion coefficient, inventories, leaching test, permeability measurement, gas measurements, radiochemistry, imaging. In conclusion, CHICADE is a nuclear facility with unique equipment, allowing exhaustive expertise to be carried out in a single location, benefiting from cross and complementary methods.

STORAGE: A Source Term Model for Intermediate-Level Radioactive Waste

The French Alternative Energies and Atomic Energy Commission and Orano have developed a modeling tool named the Simulation TOol Of RAdiolysis Gas Emission (STORAGE) for assessing gas generation of intermediate-level waste (ILW). The first version of this model was designed to estimate gas (more specifically hydrogen) production by radiolysis of organic and water-containing materials. The code deals with different types of waste packages: metal drums, concrete drums, bitumen packages, and compacted waste containers. Diverse radioactive waste can be handled: spent fuel cladding, reprocessing sludge, contaminated technological waste (gloves, bags, bottles, etc.), ion exchange resin, etc. The validity of the model was evaluated using a series of measurements performed on U,Pu–contaminated solid waste from the Orano plutonium laboratories at the MELOX facility. A benchmark study for compacted waste containers was also implemented; the results of STORAGE were compared with reference calculations performed by Orano Projets. Future improvements of the STORAGE model are also presented.

Review of CALORRE Calorimeter Characterizations Under Laboratory and Irradiation Conditions

This article reviews the work to date on the CALOrimeter with Radial thermal transfers for nuclear REactors (CALORRE) differential calorimeter patented by Aix-Marseille University (AMU) and the French Alternative Energies and Atomic Energy Commission (CEA) in 2015. The article presents the results obtained with the first prototype of the CALORRE calorimeter qualified under real conditions during an irradiation campaign in the MARIA reactor in 2015, including previously unpublished details. Then, studies of different CALORRE calorimetric cells characterized by experiments under laboratory conditions are described. Several configurations were studied to determine the influence of cell height, horizontal fin geometry, and structural material composition on calorimeter response. These calculations provide for a calibration protocol by generating a heat source inside each cell, with evaluation of linearity, sensitivity, range, reproducibility, response time, and absolute temperatures. Finally, within the framework of a new research program called Compact-CALORimeter Irradiations inside the MIT research reactor (CALOR-I) and financed by AMU Foundation (A*Midex), a design optimization of the calorimeter assembly was carried out in order to remove contact thermal resistances and provide a new very compact CALORRE calorimeter suited for the in-core water loop of the Massachusetts Institute of Technology (MIT) reactor (2 <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\text{W}\cdot \text{g}^{-1}$ </tex-math></inline-formula> peak nuclear heating rate). The response of this new very compact calorimeter is estimated using 3-D numerical thermal simulations under real conditions.

“Quantum memory” quantum computers and noiseless phton echoes

One of the most important applications of quantum computing is to crack classical cryptosystem. Previous studies showed that the number of qubits required to crack the widely used 2048-bit RSA cipher is about 20 million, which is far beyond the current technology for quantum computing. Recently, É. Gouzien and N. Sangouard of the French Alternative Energies and Atomic Energy Commission proposed a quantum computing architecture based on a two-dimensional grid of superconducting qubits and a three-dimensional multimode quantum memory. They showed that only 13k qubits are required to crack a 2048-bit RSA integer with the help of a long-lived quantum memory with 28 million spatial modes and 45 temporal modes. Their results clearly demonstrate the values of quantum memories in quantum computing and provide an alternative approach for building practically useful quantum computers. Quantum computers require quantum memories to work at microwave band, which remains an outstanding challenge. Based on a detailed analysis of atomic radiations during the quantum storage process, we recently proposed a noiseless-photon-echo protocol which can successfully eliminate the spontaneous emission noise in photon echoes. This protocol is expected to further enable microwave quantum storage and the construction of “quantum memory” quantum computers.

French cinema vs the bomb: Atomic science and a war of images circa 1950

This article examines the struggle over the meaning of atomic science and nuclear power in mid-century French film culture, including popular and art cinema, newsreels, television, and industrial films produced by the French atomic energy commission. It considers the nuclear as a representational problem, rooted in the enduring iconicity of the bombings of Hiroshima and Nagasaki and the difficulty of making atomic science’s value visible. Building on recent criticism of France’s ‘Trente glorieuses,’ it focuses on the less-than-glorious image of nuclear power that filled French screens during the 1940s and 1950s, especially in two films released in 1950: René Clair’s popular but largely forgotten, La beauté du diable, a Faust adaptation-cum-nuclear allegory, and Nicole Védrès’s seldom-seen (but recently restored and re-released) La vie commence demain (the follow-up to her better-known Paris 1900 (1947)), which uses a who’s who of French arts and letters—including Sartre, Gide, Prévert, Picasso, Le Corbusier and Jean Rostand, among others—to denounce nuclear weapons while imagining a future enlightened by peaceful atomic science.

Modular implementation framework of partitioned path-following strategies: Formulation, algorithms and application to the finite element software Cast3M

Damage, cracking, and strain localization mechanisms often lead to unstable structural responses characterized by snap-backs (i.e., force and displacement decrease simultaneously). Standard nonlinear Newton-based solution algorithms with displacement/force control cannot capture the equilibrium curve in its entirety. This can be overcome using path-following formulations. A general (i.e., valid for any finite element code) implementation framework can be designed and applied to the Cast3M software by collecting the essential concepts and the formalism of partitioned path-following arc-length algorithms. Thanks to these developments, Cast3M is now capable of processing path-following equations without any major modifications. Three path-following constraints were selected to demonstrate the applicability of this framework: a first one on the combination of the displacement increment at a given set of nodes, a second one on the maximum strain increment over the computational domain, and a third one on the maximum elastic predictor of the damage/plastic criterion function over the computational domain. Two- and three-dimensional strain localization simulations show that the proposed framework behaves in a stable and convergent manner, even when multiple severe snap-back instabilities are present. Users of Cast3M shall find the proposed study helpful in that it allows them to focus on developing new path-following equations for the software. Cast3M is developed by the French Alternative Energies and Atomic Energy Commission (CEA) and freely available for research purposes. The developments discussed in this paper have been made available to the user/developer community along with Cast3M 2021 (release date: June 2021).

Debonding of rubberised fibre-reinforced cement-based repairs under fatigue loading: experimental study and numerical modelling

ABSTRACT This study presents experimental and numerical studies of the initiation and propagation of interface debonding between the substrate and thin-bonded cement-based overlay under fatigue testing. The overlay material is mortar incorporating amorphous metallic fibres (30 kg/m) and/or rubber aggregates (30%), while the substrate is a plain mortar. Uniaxial tensile tests were carried out on overlay materials and substrate-overlay interface to obtain a residual normal stress-crack opening relationship (σ-w law) and residual normal stress and substrate-overlay interface debonding relation respectively. Three-point bending fatigue tests were carried out on the composite beam to analyse the structural behaviour of repaired beam. The propagation of debonding along the substrate-overlay interface was monitored by using the Digital 3D Image Correlation (DIC) technique. Three-point bending fatigue test was modelled by finite element method (FEM) using CAST3M, developed by French Atomic Energy Commission. Direct tension test results show reduction in tensile strength and improvement in strain capacity (1.5 times of control mortar) for rubberised mortars. For fibre-reinforced mortars, post-peak residual tensile strength is significantly enhanced. Rubberised fibre-reinforced mortars show positive synergetic effects both in the improvement of residual post-peak strength and strain capacity (3.5 times than control mortar). Results show that the rubberised fibre-reinforced thin-bonded cement-based overlay proved to be very effective in restraining crack opening, by transferring stresses through the crack, and enhancing the fatigue resistance of repair system, by limiting interface debonding propagation upto 50%. Experimental results of composite beams are in good agreement with modelling results.

Bayesian Approach for Multigamma Radionuclide Quantification Applied on Weakly Attenuating Nuclear Waste Drums

Gamma spectrometry is a passive nondestructive assay method used to quantify radionuclides present in nuclear objects. Basic methods using empirical calibration with a standard to quantify the activity of nuclear materials by determining the calibration coefficient are ineffective on nonreproducible nuclear objects such as waste packages. Package specifications such as composition or geometry change from one package to another and exhibit large variability of objects. The current standard quantification process uses numerical modeling of the measured scene with few available data such as geometry or composition, in particular density, material, screen, geometric shape, matrix composition, matrix, and source distribution. Some of them are strongly dependent on package data knowledge and operator backgrounds. The French Atomic Energy Commission (CEA) is developing a methodology to quantify nuclear materials in waste packages and waste drums without operator adjustment and internal package configuration knowledge. This method suggests combining a stochastic approach which uses, among others, surrogate models available to simulate the gamma attenuation behavior, a Bayesian approach considering conditional probability densities and prior information of problem inputs, and Markov Chain Monte Carlo (MCMC) algorithms which solve inverse problems, with gamma ray emission radionuclide spectra, and the outside dimensions of the objects of interest. The methodology has been tested to quantify actinide activity with a low bulk density matrix, weakly attenuating compositions, without information on the distribution of the source in terms of actinide masses and materials composing the drums. Activity uncertainties are taken into account.

Thermo-mechanical behavior of an ablated reactor pressure vessel wall in a Nordic BWR under in-vessel core melt retention

The reactor pressure vessel (RPV) of a nuclear reactor is one of the key safety barriers preventing radioactive environmental releases during a severe accident. One of the promising strategies of severe accident management (SAM) is to retain the molten core having continuous decay heat inside the RPV by natural water cooling of the external vessel surface. The feasibility of such a strategy relies on complex safety analyses including accurate prediction of vessel thermo-mechanical behavior which can be assessed by mechanical stresses and strains.In this paper, we present the stress–strain response of an ablated RPV of a Nordic boiling water reactor (BWR) to dynamic thermomechanical loads set by expanding volumetrically heated molten pool inside the RPV cooled by water at the external surface. MELCOR 2.2.9541 severe accident code is used to simulate the in-vessel behavior and provides the input conditions for dedicated structural analysis of the RPV using ANSYS® Mechanical APDL 19.2. A creep model of the SA533B1 vessel steel is validated against uniaxial creep tests carried out by INEL (Idaho National Engineering Laboratory) and creep tests performed at CEA (French Alternative Energies and Atomic Energy Commission) as part of the OLHF (OECD Lower Head Failure) Project. Two generic severe accident scenarios are considered: (i) Station Blackout (SBO) and (ii) Station Black-out and Loss-of-coolant Accident (SBO + LOCA). In both scenarios, we found that the RPV has maintained structural integrity considering two failure criteria: stress-based and strain-based.

Comparison of Results from Recent NNSA and CEA Measurements of the 239Pu(n, f) Prompt Fission Neutron Spectrum

The National Nuclear Security Administration (NNSA)/DP French Alternative Energies and Atomic Energy Commission (CEA)/DAM agreement on cooperation on fundamental science is a U.S.-French collaborative effort to combine intellectual and experimental resources and further the relevant nuclear science. Recently, both the NNSA and CEA experimental teams performed high-statistics measurements of the 239Pu(n, f) prompt fission neutron spectrum (PFNS) at the Los Alamos Neutron Science Center, both of which were recently published in the journal Physical Review C. These separate measurements used the same experimental area and a common neutron detector array, but differ in many aspects, including background assessments, data acquisition systems and philosophies, fission detectors, and PFNS extraction techniques. Hence, some aspects of the experimental methods and associated uncertainties are highly correlated while others are independent. The results from both measurements broke new ground for PFNS measurements given their higher accuracy and more detailed study of corrections necessary for the measured quantity compared to existing literature measurements, and both will significantly impact PFNS nuclear data evaluations for the foreseeable future. The focus of this work is to document a comparison of the results from these distinct measurements in terms of the acquired data, the PFNS results, and the measured average PFNS energies. While systematic differences between the PFNS results are present on the 1–3% level, the acquired data relative to each respective measurement at low incident neutron energies are in remarkable agreement, as are the conclusions regarding the magnitude and position of features in the PFNS relating to second-chance fission, third-chance fission, and pre-equilibrium neutron emission.

Pulse-reactor core monitoring with an innovative optical neutron detector

To minimise uncertainty and strengthen estimation of energy deposited during power transients, we set up an online passive neutron flux measurement of a pulse reactor.Optical fission chambers, are being developed in the framework of dependable neutron flux instrumentation for Sodium-cooled Fast Reactor (SFR) and Molten-Salt Reactor (MSR) of Generation IV by the French Atomic Energy and Alternative Energies Commission (CEA). Neutron interactions with a fissile layer coating release heavy-ions in the MeV-range, eventually leading to spontaneous photon emission in the ultraviolet to infrared range of the surrounding noble gas. Prototypes have previously been tested in an ideal pure neutron field. In this paper, linearity of optical signal versus reactor power is assessed during startups and high power transients up to 21.6 GW. Excellent linearity results are obtained over 5 reactor power decades and detection remain possible over 7 decades. Detection range is only restrained by photodetector properties. Such detector may be routinely employed for high-dynamic transients monitoring by the core caisson of pulse reactors.