The European Journal of Physics N (EPJ-N)
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Has natural radioactivity contributed to the evolution of living organisms? Validation of a dedicated calculation scheme (for isotopic concentrations and deposited energies) on Oklo's natural nuclear reactors
No full textIn October 2018, the French Society of Nuclear Energy (SFEN/PACA) organized a conference-debate entitled “What happened two billion years ago where present-day Gabon is located? Has Nuclear Power contributed to the development of Life on Earth?” Our contribution to this reflection concerns the possible impact of radiation emitted from radioactive media on the early evolution of living organisms and may also be useful for research on the storage of radioactive waste. This article concerns the qualification of a calculation scheme for isotopic concentrations and energy deposits in Oklo's natural nuclear reactors or any other naturally radioactive environment. Our work has been carried out using a calculation scheme involving the Monte Carlo software TRIPOLI-4® [TRIPOLI-4 R, CEA, EDF and AREVA reference Monte Carlo code, https://doi.org/10.1016/j.anucene.201
Looking ahead to severe accident research
No full textSevere Accident (SA) research is currently facing new challenges coming from changes in the energy and computer science sectors. Therefore, it is imperative to reassess the current status in the area to optimize where the research resources should go to reach even higher safety standards both in the running Nuclear Power Plants (NPP) and the upcoming new designs, particularly Water-Cooled Small Modular Reactors (WC-SMR). Three Horizon Euratom projects stand out in such context. SEAKNOT (SEvere Accident research and KNOwledge managemenT) is progressing in setting a SA research roadmap by ranking the major phenomena involved in terms of knowledge and safety significance in LWRs (large water-cooled reactors) and SMRs, at the same time that it is strengthening paths for Education & Training (E&T) on SA for forthcoming generations of researchers and engineers. SASPAM-SA (Safety Analysis of SMR with PAssive Mitigation strategies-Severe Accident) is supplying valuable information on phenomena, boundary and accident conditions that might prevail in WC-SMRs, specifically integral PWR (iPWR). The project allows the assessment of the applicability of the current state-of-the-art simulation codes and the relevance of large reactor experiments to iPWRs. Different SA mitigation strategies, like In-Vessel Melt Retention (IVMR), are being explored. Finally, ASSAS (Artificial intelligence for Simulation of Severe AccidentS) is working to prove the possibility to develop fast-running SA simulators thanks to Artificial Intelligence, to support training, engineering and emergency response. This paper discusses the major progress made in the three projects and their complementarity contributes to a safer nuclear energy production
Sensitivity and uncertainty calculations in support of the noise analysis of the TAPIRO reactor
No full textIn the framework of an assessment of the application of neutron noise methodologies to the TAPIRO reactor, sensitivity and uncertainty analyses on impacting nuclear data for this experimental reactor have been performed using the Serpent Monte Carlo code. Integral parameters relevant to the noise analysis of TAPIRO have been investigated, namely the effective multiplication factor, the effective delayed neutron fraction, the effective generation time and the prompt neutron decay constant. The nuclear data uncertainty has been assessed using different data libraries (ENDF/B-VIII.0 and JEFF-3.3), different energy grids (ECCO-33 and XMAS-172) and different techniques (Generalised Perturbation Theory and Unscented Transform). A ranking of the most impacting nuclides (235U, 63Cu, 65Cu and 238U) and reactions (elastic scattering, fission, radiative capture and fission neutron yield) in terms of uncertainties has been performed, pointing out that the impact of nuclear data uncertainties on the responses is non-negligible and must be taken into account when the noise analysis of the reactor is performed. In particular, the uncertainty on the value of the prompt neutron decay constant due to effect of the nuclear data uncertainties is larger than 100%, especially when the system under study is close to criticality
Advanced reactor models for plutonium multirecycling scenario studies using artificial neural networks
No full textThis paper presents an innovative way to identify the plutonium content in MOX fresh fuel according to its isotopic composition in fuel cycle dynamic simulations. In those calculations, the fuel fabrication is a challenge to model as the plutonium stock status (quantity and quality) is unknown and the fuels that are loaded in reactors should ensure criticality throughout the wanted cycle length. This model is based on metamodeling full core depletion simulation of PWRs with artificial neural networks to predict efficiently the cycle length and power factors as a function of the MOX fresh fuel composition. An exploration of the plutonium phase space has shown the impact of plutonium isotopic composition on power factors and fuel cycle length variability. Moreover, the sensitivity of the cycle length to the plutonium content have led us to propose a model that links the plutonium content to its isotopic composition by a constant step function, while accepting an uncertainty of the order of 1 GWd/t on the cycle length. The impact of this new model on PWR loaded with plutonium recovered by UOX and MOX spent fuel reprocessing is assessed in this paper. Pu content estimator based on this new approach leads to standard deviation of the cycle length discrepancies distribution lower than 150 MWd/t for a Pu fissile quality range size of 4%
Modeling of fuel salt compressibility in MSR using blastFOAM, reference and accelerated calculations using transfer function
No full textThis article presents the work performed to compute the impact of compressibility on the fuel salt of Molten Salt Reactors (MSRs) on reactivity density feedback evolution in the OpenFOAM solver, blastFoam. Reactivity density feedback is a crucial aspect of the chain reaction stability in MSRs. The delay induced by the compressibility of the salt is particularly significant to simulate, especially during transient power variations of very short duration. This study aims to describe the delay induced by the compressibility of the fuel salt in the evolution of density neutronic feedback. To perform this computation, the Correlated Sampling (CS) method was used to compute the neutronic feedback and chained to the compressible results. Based on the results of the study, a methodology is described for obtaining a transfer function to compute density neutronic feedback with only one compressible calculation. This transfer function is computed for verification and illustration purpose on a simplified reactor model and the Molten Salt Fast Reactor (MSFR). It would enable future work to incorporate this physics into non-compressible codes (such as CFD codes and system codes) with minimal computation time
Global Evaluated Nuclear File (GENF/D): a decay data sub-library for radionuclide inventory applications
No full textDecay data properties are regularly evaluated and re-evaluated, more frequently than are published evaluated decay nuclear sub-library. This work presents a procedure for generating an up-to-date decay library in ENDF-6 format, based on NUBASE2020, supplemented with the latest ENSDF data, through IAEA Livechart API, and extended with BetaShape, BrIcc, RadiationReport and GEF predictions. Called Global Evaluated Nuclear File (GENF/D), this decay sub-library covers 5502 isotopes, including 3511 ground states, and describes 8531 decay channels. Special care has been taken for spontaneous fission description, and using 2018 ENDF-6 format update, explicits the neutrino and anti-neutrino spectra. The former task led to the production of a new version of GEFY database for spontaneous and neutron-induced fission fragments database, based on NUBASE2020. Comparisons to the latest distributions of JENDL, ENDF/B and JEFF have been made, as well as as comparison to benchmarks of decay heat from shutdown reactors, with fissile and non-fissile materials
COG 11.3 New Features
No full textCOG is a high fidelity, multi-particle, Monte Carlo radiation transport code in development at LLNL since 1980 in support of multiple applications including nuclear criticality safety, radiation shielding, radiography, and subcritical multiplicity analysis. COG 11.3 is the latest version scheduled for public release in 2024 representing the culmination of ten years of development since previous releases. This paper describes the major new features available in COG 11.3 including: (a) updated nuclear data libraries; (b) updated activation data; (c) advanced LLNL Fission Reaction Event Yield Algorithm; (d) new time-tagged list-mode detector feature; (e) new spontaneous fission source feature; (f) new alpha and deuteron particle transport; (g) updated electron transport (implementing EGS5 with internal electron library generation using PEGS5); (h) three new estimators for the effective delayed neutron fraction; (i) new input pre-processor options for user-friendly generation of three-dimensional rectangular and triangular lattice geometries; (j) new parallel-processing capabilities for the inverse reactor period and CritDetVR features using MPI; (k) a new imaging detector feature; and (l) details of the modernized COG website at https://cog.llnl.go
The PIANOFORTE partnership: Elevating European research for enhanced radiation protection
No full textThe PIANOFORTE partnership (2022–2029) aims to enhance radiation protection for the public, patients, workers, and the environment across various exposure scenarios. This European initiative addresses key barriers in health and environmental risk research related to ionising radiation and promotes findings that support effective radiation protection policies. By building a comprehensive pan-European scientific and technological foundation, PIANOFORTE ensures that the radiation protection system remains fit-for-purpose, delivers science-based policy recommendations and improved practices across sectors using nuclear technology and ionising radiation, including both energy-related and non-energy applications. In the medical field, PIANOFORTE works to reduce uncertainties in health risk estimates and support innovations in cancer diagnosis and therapies. Other key priorities include developing reliable methods for evaluating radiation protection related to new technologies and managing radiation emergencies, improving strategies for both immediate response and long-term recovery. The Partnership's multi-stage prioritisation mechanism of research needs ensures that developed efforts reflect the perspectives of a broad range of stakeholders, including researchers, policy makers, regulators, implementers and practitioners. This inclusiveness aligns research priorities with pressing societal challenges, such as climate change impacts and nuclear technology safety. PIANOFORTE's open call process funds research projects that align with its strategic goals, expanding its network from 58 to 108 partners after inclusion of new partners of granted projects during the two first open calls. Additional calls will continue to foster collaboration and increase research capacity across Europe. By adopting FAIR (Findable, Accessible, Interoperable, and Reusable) data management practices and embracing open science, PIANOFORTE supports the broader radiation protection community in sharing infrastructure and research outcomes. Educational initiatives are central to PIANOFORTE's mission, as it builds Europe's expertise in radiation protection through training programmes for current and next generation scientists. Structured dialogue with stakeholders strengthens the Partnership's impact, bridging research and policy and helping to create a well-informed, resilient society capable of making sound, risk-aware decisions about nuclear and radiation-related issues
Multi-physics DONJON5 reactor models for improved fuel cycle simulation with CLASS
No full textThis work investigates reactor model biases and their consequences in nuclear scenario simulations. Usually, the models for Pressurized Water Reactors are based on infinite 2D assembly depletion simulations, but recent work has shown the importance of 3D complete core simulation for uncertainty reduction. The consideration of a whole core leads to new reactor parameters in the simulations that may bring additional biases. The fuel temperature distribution is one of them, and previous work considered isothermal reactors, leading to probable uncertainties in spent fuel inventory at reactor discharge. To quantify those biases and their propagation in a full scenario simulation, new advanced reactor models have been developed, based on neutronics and thermal-hydraulics couplings at the core level performed with DONJON5. Results show that the plutonium isotopic quality of spent fuel is biased for an isothermal core, with values systematically higher than for multi-physics calculations. In order to propagate those discrepancies in fuel cycle simulations that involve plutonium recycling in PWR MOX fuels, the coupling between CLASS and DONJON was renewed in order to add new fuel parameters such as the fuel temperature in the core burn-up simulation. A new methodology for data interpolation from lattice calculation has been implemented that allows acceptable computational time for DONJON5 calculations that are done within the fuel cycle simulation performed by CLASS. Comparison between isothermal and multi-physics reactor models for advanced scenario simulations performed with CLASS shows that the isothermal hypothesis leads to biases up to 10% for plutonium inventory in the UOX spent fuel stockpile, comparable with biases associated with other reactor parameters such as the loading pattern
Data assimilation of decay heat experiments for feedback on nuclear data
No full textIntegral decay heat experiments can provide interesting feedback on particular nuclear data (decay data and fission yields mainly). After ensuring that the C/E discrepancies were mostly due to nuclear data discrepancies, a Bayesian inference approach can be applied. Nevertheless, the results strongly depend on the quality of the experiment and on our capability to estimate realistic experimental correlation matrices when considering several integral experiments in the assimilation process. A former study performed in 2019 was dedicated to the data assimilation of a large C/E dataset from the experimental validation database of fuel inventory calculations with the DARWIN2.3 package in order to provide feedback to the nuclear data evaluators. This paper is an attempt to exploit the General Electric decay heat experiments performed in the USA in the 1980s in order to confirm or not the trends on four particular cumulated fission yields: 235U(nth,f)133Cs, 235U(nth,f)137Cs, 239Pu(nth,f)106Ru and 239Pu(nth,f)144Ce