The European Journal of Physics N (EPJ-N)
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Increase of nuclear installations safety by better understanding of materials performance and new testing techniques development (MEACTOS, INCEFA-SCALE, and FRACTESUS H2020 projects)
No full textResearch to better understand the phenomena influencing materials and components’ performance is important for increasing the safety of Generation II and III nuclear plants. A crucial step for improving nuclear safety is the development of new experimental techniques that can provide the necessary data. The three H2020 projects presented in this paper, MEACTOS (2017–2022), INCEFA-SCALE (2020–2025), and FRACTESUS (2020–2024), cover the steps needed to realize those safety improvements. The goal of the MEACTOS project is to improve the resistance of critical locations, including welds, to environmentally-assisted cracking through optimizing surface machining and treatments. The project is currently in its final stage, and the complete analysis of the data is finished. The objective of INCEFA-SCALE is to improve predictions of component fatigue lifetime when subjected to Environmentally-Assisted Fatigue (EAF). The strategy consists of producing guidance on how to appropriately accommodate variable amplitude and plant-relevant loading in EAF assessments. Increasing the understanding of the EAF mechanism based on substantial testing, characterization, and analysis program will support the INCEFA-SCALE strategy. The FRACTESUS project will validate the use of miniaturized compact tension specimens by comparing the results of master curve-oriented fracture toughness tests performed with small and large specimens. The round-robin exercises will use irradiated and non-irradiated Reactor Pressure Vessel (RPV) materials. The material selection process is complete in time for the project to enter the testing phase. The output of the project will be beneficial from a long-term operation perspective and a saving in the material amount needed for RPV surveillance programs. Even though each project is devoted to different research areas, common aspects are clearly visible. All three projects investigate phenomena that are relevant to the performance and safe operation of the nuclear plant. Moreover, each project will provide valuable databases and analyses of test results for materials relevant to components in the nuclear plant. The output of these projects will be of great value to the nuclear industry. This paper presents the current progress for each project, emphasizing the common research domains between the projects
DONJON5/CLASS coupled simulations of MOX/UO
No full textMost fuel cycle simulation tools are based either on fixed recipes or assembly calculations for reactor modeling. Due to the high number of calculations and extensive computational power requirements, full-core computations are often seen as not viable for this purpose. However, this leads to additional hypotheses and modeling biases, thus limiting the realism of the resulting fuel cycle. For several applications, the current modeling method is sufficient, but precise calculations of discharged fuel composition may require further refinements. CLASS (Core Library for Advanced Simulation Scenarios) is a dynamic fuel cycle simulation code developed since 2012 with reactor models based on neural networks to produce nuclear data and physical quantities. Past work has shown a first coupling between CLASS and DONJON5 to quantify neural networks approach biases. This work assesses the applicability of 3D full-core diffusion calculations using the DONJON5 code coupled with nuclear scenario simulations involving a realistic PWR core at equilibrium cycle conditions. DONJON5 interpolates burnup dependent diffusion coefficients and cross sections generated beforehand by DRAGON5, a deterministic lattice calculation tool. Whereas previous studies considered only homogeneous reactors (i.e. homogeneous assembly in terms of composition and enrichment as well as homogeneous core), the present contribution focuses on the integration of full-core calculations in CLASS for fuel cycles involving a MOX/UO2 PWR core (i.e. 1/3 MOx–2/3 UOx). The DONJON5 model considered in this work describes a core with critical boron concentration at each time step partially loaded with MOx heterogeneous assemblies composed of three enrichments. In fuel cycle calculations, the main issue is to adapt, in the fabrication stage, the fresh fuel composition for the reactor with regards to the isotopic composition of the available stocks. This work presents a fuel loading model based on power peaking factors minimization that respects irradiation cycle length, 235U enrichment as well as Pu concentration and fissile quality, hence, ensuring a more uniform power distribution in the core
Medical applications of ionizing radiation and radiation protection for European patients, population and environment
No full textMedical applications of ionising radiation (IR) represent a key component of the diagnosis and treatment of many diseases, guaranteeing efficient health care. The use of IR in medicine, the largest source of general population radiation exposure, is potentially associated with increased risk of cancer and non-cancer diseases, which needs to be evaluated to provide evidence-based input for risk-benefit considerations. Efforts are also needed to improve the safety and efficacy of medical applications through optimisation. The EC Euratom programme enhances research in medical radiation protection. The four complementary multidisciplinary projects presented here contribute to (1) improving knowledge on exposure and effects of diagnostic and therapeutic applications and (2) transferring results into clinical practice. The common aim is to optimise use for individual patients, enhance education and training, ensuring adherence to ethical standards, particularly related to technologies based on artificial intelligence. MEDIRAD, SINFONIA and HARMONIC focus on improving exposure estimation and studying the detrimental effects of diagnostic and therapeutic medical exposures in patients and staff using different endpoints. EURAMED rocc-n-roll brings together the results of the projects and the recommendations generated by them to build, in collaboration with the EU Radiation Protection research platforms, a strategic research agenda and a roadmap for research priorities
In-Can vitrification of ALPS slurries from Fukushima Daiichi effluent treatment using DEM&MELT technology
No full textAfter the accident at the Fukushima Dai-ichi Nuclear Power Station, a large amount of contaminated water was treated using several decontamination systems with different natures of adsorbents and chemicals. The resulting wastes, called Fukushima Effluent Treatment Wastes (FETW), were stored at the Fukushima Dai-ichi site. Vitrification could be the most promising treatment method to package these wastes. The consortium gathering CEA, Orano, ECM Technologies and ANDRA, implemented an in situ, robust, simple and versatile In-Can vitrification process, the DEM&MELT technology. Since 2018, the applicability of this technology for FETW treatment and conditioning has been evaluated. In 2021–2022, studies focused on one particular waste, coming from the ALPS system (Advanced Liquid Processing System-Multi Radionuclides Removal) generating around 70%vol. of FETW. This waste is composed of two co-precipitation slurries: one mainly composed of iron hydroxide, and one of calcium carbonate and magnesium hydroxide. The purpose of this article is to highlight the feasibility of ALPS slurries vitrification with DEM&MELT, relying on tests performed from laboratory-scale to full-scale. Macroscopically homogeneous glasses were produced using the DEM&MELT demonstrator, with a waste loading of 60 wt.% (expressed as waste dry mass) and microstructural analyses were performed. It gives promising results for FETW conditioning with the DEM&MELT process
Impact of H in H
No full textIn this paper, the impact of the thermal scattering data for H in H20 is estimated on criticality benchmarks, based on the variations of the CAB model parameters. The Total Monte Carlo method for uncertainty propagation is applied for 63 keff criticality cases, sensitive to H in H20. It is found that their impact is of a few tenth of pcm, up to 300 pcm maximum, and showing highly non-linear distributions. In a second step, an adjustment is proposed for these thermal scattering data, leading to a better agreement between calculated and experimental keff values, following an increase of scattering contribution. This work falls into the global approach of combining advanced theoretical modelling of nuclear data, followed by possible adjustment in order to improve the performances of a nuclear data library
Core and safety design for France–Japan common concept on sodium-cooled fast reactor
No full textFrance (CEA and FRAMATOME) and Japan (JAEA, MFBR, and MHI) teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor (SFR) concept. This paper mainly describes the capabilities of ASTRID 600 to demonstrate SFR technologies of both countries. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600, considering Japan’s target for core performance. The neutronics design of the core has satisfied most required design targets and conditions. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device (CSD), called a self-actuated shutdown system (SASS), one of the safest approaches in Japan. Japan’s team used the SASS in place of the hydraulically suspended absorber rod, called RBH, one of the safest approaches of France, to investigate the potential of the SASS as a design measure common to the countries. The preliminary analysis has shown that the SASS can satisfy the countries’ main requirements. This study has also revealed that the mitigation measures of ASTRID 600 against a severe accident are promising to achieve in-vessel retention for both countries
Diluent effects on the stability range of w/o micellar systems and microemulsions made with anionic extractants
No full textHere we present a series of complete phase prisms for water, an organic diluent and di-(2-ethylhexyl) phosphoric acid (HDEHP), one of the most widely used double-branched lipophilic surfactants in hydrometallurgy. Partial or total titration with sodium hydroxide evidence that the mole fraction of the counter-cation “Z” is the variable that controls the packing and spontaneous curvature of the curved film formed by this extractant. Penetrating solvents such as toluene and iso-octane and the non-penetrating solvent dodecane as well as common hydrotropes acting as co-solvents, are considered. The three classical cuts of the phase prism are shown. The regions for which liquid–liquid extraction is possible are determined, as well as the location of the liquid crystals at the origin of the often observed third-phase formation. It is shown that profoundly different trends are obtained when replacing the common solvents currently used in hydrometallurgical processes with hydrotropes
Investigations on the source term of the detection of radionuclides in North of Europe in June 2020
No full textDuring the second half of June 2020, small quantities of artificial radionuclides (60Co, 134Cs, 137Cs, 103Ru, 106Ru, 141Ce, 95Nb, 95Zr) have been detected in northern Europe (Finland, Sweden, Estonia), the source of the release being unknown. The measured values were close to detection limits and didn’t represent any health issue. This paper presents the investigations carried out at IRSN in order to identify the release origin. The most probable source location and the release magnitude estimation are briefly presented. This recent set of detection is also compared to previous similar ones. This paper mainly focuses on the investigations which have been performed in order to answer two main questions. First “from which type and part of a nuclear installation the release could come from?”. Although no certainty is achievable, the most probable source is found to be a spent primary ion exchange resin. The second question addressed was “how this radiological inventory could have been released into the atmosphere?”. But, mainly due to the lack of information, no satisfying answer has been found to that question and what really happened remains unknown
Long-term radionuclide retention in the near field: collaborative R&D studies within EURAD focusing on container optimisation, mobility, mechanisms and monitoring
No full textWithin EURAD, targeted collaborative research activities are performed to further deepen understanding regarding the long-term behaviour of key components in the repository near-field, assess specific radionuclide retention processes as well as developing methods for monitoring safety relevant parameters of repository systems. The ambition of the four EURAD Workpackages (WPs) – CONCORD, FUTURE, CORI, MODATS – presented here, is to investigate topics to meet implementation needs and contribute to Safety Cases in Europe at the highest level of scientific excellence. Work is fully integrated into the EURAD concept, emphasizing interactions between different WPs, involvement of End Users, assuring the link to national programmes and contributing to overarching features like Knowledge Management, Training and Education, or European Integration. Comprehensive initial State-of-the-Art reports were prepared by the WPs or currently under development and are available at the EURAD website. The technical/scientific work performed in the four WPs - CONCORD, FUTURE, CORI, MODATS – is discussed in this contribution
Algorithms for processing self-powered neutron detector signals important for determination of local parameters in each part of the VVER core
No full textThese investigation findings prove the possibility of a engineering solution for VVER core automatic protection during operation in both nominal and transient conditions within ICIS using local parameters (i.e. linear heat power of the most stressed fuel rod, departure from nucleate boiling ratio). Such engineering solution will be implemented by safety system software-hardware (PTK-Z) on the basis of signals coming from in-core neutron flux detectors, temperature sensors, primary coolant flow and coolant pressure transducers. The article presents the following: a list of in-core neutron flux detectors, a list of transducers of primary coolant monitoring for thermal engineering conditions (temperature, flow rate, pressure), whose signals are delivered to the terminals of instrumentation gauges (PTK-Z) for the purpose of actuating safety actions in accordance with local parameters. The paper shows arrangement of in-core neutron flux detector equipment, above view of self-powered neutron detector (SPND) location in safety channels, axial arrangement of SPND along the core, algorithms for processing SPND signals important for determination of local parameters in each part of the core in both normal and abnormal operation conditions