The European Journal of Physics N (EPJ-N)
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A journey into Massimo Salvatores scientific work
No full textMassimo SALVATORES was not the man of a country, an organization, or a team. Certainly because of his origin, his education, and his culture, Massimo has always favored a broader and more open collaboration instead of a bureaucratic and shortsighted approach to the research, keeping the achievements only to a restricted inner circle. He was convinced that disinterested sharing makes one stronger and Massimo is one of the few nuclear reactor physicists who elevated international collaboration to its highest level. A short history of the major contributions that Massimo made to his dear discipline, Neutronics, will emphasize this peculiar aspect of his career
NMB4.0: development of integrated nuclear fuel cycle simulator from the front to back-end
No full textNuclear Material Balance code version 4.0 (NMB4.0) has been developed through collaborative R&D between TokyoTech&JAEA. Conventional nuclear fuel cycle simulation codes mainly analyze actinides and are specialized for front-end mass balance analysis. However, quantitative back-end simulation has recently become necessary for considering R&D strategies and sustainable nuclear energy utilization. Therefore, NMB4.0 was developed to realize the integrated nuclear fuel cycle simulation from front- to back-end. There are three technical features in NMB4.0: 179 nuclides are tracked, more than any other code, throughout the nuclear fuel cycle; the Okamura explicit method is implemented, which contributes to reducing the numerical cost while maintaining the accuracy of depletion calculations on nuclides with a shorter half-life; and flexibility of back-end simulation is achieved. The main objective of this paper is to show the newly developed functions, made for integrated back-end simulation, and verify NMB4.0 through a benchmark study to show the computational performance
Estimation of the vitrified canister production for a PWR fleet with the CLASS code
No full textThis article presents an assessment of fuel cycle parameter impact on waste production through the prism of vitrified container and minor actinide masses, using a scenario simulated with the CLASS code. The number of canister introduces a new focus on waste production estimation for a nuclear fleet, as it helps to set the repository size for deep geological disposal of high level waste. To evaluate the number of canisters, dedicated developments to model a simplified waste vitrification unit in the CLASS package are presented. It relies on artificial neural network estimations of decay heat, α radiation and mass content, for different material flow coming from reprocessing and sent to vitrification. Then, the studied scenario considers a transition from a PWRs plutonium mono-recycling fleet to a plutonium multi-recycling fleet. Vitrified waste container production is calculated as a function of different material reprocessing options. Simulations shows that up to 19% variation may be observed (in 2060) in canisters’ total number depending on the different assumptions. Impact of vitrification parameters such as the size of buffer before vitrification is also analysed and the importance of mixing material coming from MOX and MIX spent fuels with material from UOX spent fuels is clearly established
Impact of H in H
No full textThe impact of the H in H2O thermal scattering data are calculated for burnup quantities, considering models of a UO2 pincell with DRAGON and SERPENT. The Total Monte Carlo method is applied, where the CAB model parameters are randomly varied to produce sampled (random) LEAPR input files for NJOY. A large number of burnup calculations is then performed, based on the random thermal scattering data. It is found that the impact on k∞ is relatively small (less than 35 pcm), as for nuclide inventory (less than 1% at 50 MWd/kgU) and for decay heat (less than 0.4%). It is also observed that the calculated probability density functions indicate strong non-linear effects
Delayed gamma fraction determination in the zero power reactor CROCUS
No full textGamma rays are an inextricable part of a nuclear reactor’s radiation field, and as such require characterization for dose rate estimations required for the radiation protection of personnel, material choices, and the design of nuclear facilities. Most commonplace radiation transport codes used for shielding calculations only included the prompt neutron induced component of the emitted gamma rays. The relative amount of gamma rays that are emitted from delayed processes – the delayed gamma fraction – amount to a significant contribution, e.g. in a typical zero power reactor at steady state is estimated to be roughly a third. Accurate predictions of gamma fields thus require an estimation of the delayed content in order to meaningfully contribute. As a consequence, recent code developments also include delayed gamma sources and require validation data. The CROCUS zero power research reactor at EPFL is part of the NEA IRPhE and has therefore been characterized for benchmark quality experiments. In order to provide the means for delayed gamma validation, a dedicated experimental campaign was conducted in the CROCUS reactor using its newly developed gamma detection capabilities based on scintillators. In this paper we present the experimental determination of the delayed gamma fraction in CROCUS using in-core neutron and gamma detectors in a benchmark reactor configuration. A consistent and flexibly applicable methodology on how to estimate the delayed gamma fraction in zero power reactors has hitherto not existed – we herein present a general experimental setup and analysis technique that can be applied to other facilities. We found that the build-up time of relevant short lived delayed gamma emitters is likely attributed to the activation of the aluminium cladding of the fuel. Using a CeBr3 scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6±0.6)%
The heuristically-based generalized perturbation theory
No full textThe generalized perturbation method is described relevant to ratios of bi-linear functionals of the real and adjoint neutron fluxes of critical multiplying systems. Simple linear analysis for optimization and sensitivity studies are then feasible relative to spectrum and space-dependent quantities, such as Doppler and coolant void reactivity effects in fast reactors
Pu multi-recycling scenarios towards a PWR fleet for a stabilization of spent fuel inventories in France
No full textNuclear scenario studies are performed to explore the impact of possible evolutions of nuclear fleets. The nuclear fuel cycle simulation tool COSI, developed by CEA, is used to model these dynamic scenarios and to evaluate them with respect to uranium and plutonium management, fuel reprocessing and waste production. In recent years, scenarios have focused on transitions from the current nuclear French fleet to a deployment of SFR. However, the French Multi-annual Energy Planning has recently postponed the deployment of this technology to the second half of the 21st century. Alternative solutions of plutonium management in PWR are investigated to stabilize total inventories of spent nuclear fuels. The MIX concept is based on homogeneous fuel assemblies where fuel rods are composed of plutonium blended with enriched uranium. In this study, a transition from the current French fleet to an EPR™ fleet is simulated. Two power capacities of the future EPR™ fleet are considered. A progressive deployment of fuel multi-recycling in the EPR™ fleet is implemented to enable stabilization of all spent fuels and plutonium inventories. Natural uranium consumption is also minimized thanks to ERU fuel batches in EPR™. Results are compared with plutonium and uranium mono-recycling in a PWR fleet
Sea container inspection with tagged neutrons
No full textNeutron inspection of sea-going cargo containers has been widely studied in the past 20 yr to non-intrusively detect terrorist threats, like explosives or Special Nuclear Materials (SNM), and illicit goods, like narcotics or smuggling materials. Fast 14 MeV neutrons are produced by a portable generator with the t(d, n)α fusion reaction, and tagged in both direction and time thanks to the alpha particle detection. This Associated Particle Technique (APT) allows focusing inspection on specific areas of interest in the containers, previously identified as containing suspicious items with X-ray radiographic scanners or radiation portal monitors. We describe the principle of APT for non-nuclear material identification, and for nuclear material detection, then we provide illustrations of the performances for 10 min inspections with significant quantities (kilograms) of explosives, illicit drugs, or SNM, in different cargo cover loads (e.g. metallic, organic, or ceramic matrices)
VVUQ of a thermal-hydraulic multi-scale tool on unprotected loss of flow accident in SFR reactor
No full textWithin the framework of the French 4th-generation Sodium-cooled Fast Reactor safety assessment, methodology on VVUQ (Verification, Validation, Uncertainty Quantification) is conducted to demonstrate that the CEA's thermal-hydraulic Scientific Computation Tools (SCTs) are effective and operational for design and safety studies purposes on this type of reactor. This VVUQ-based qualification is a regulatory requirement from the French Nuclear Safety Authority (NSA). In this paper, the current practice of VVUQ approach application for a SFR accidental transient is described with regard to the NSA requirements. It constitutes the first practical, progressively improvable approach. As the SCT is qualified for a given version on a given scenario, the transient related to a total unprotected station blackout has been selected. As it is a very complex multi-scale transient, the SCT MATHYS (which is a coupling of the CATHARE2 tool at system scale, TrioMC tool at component scale and TrioCFD tool at local scale) is used. This paper presents the preliminary VVUQ application to the qualification of this tool on this selected transient. In addition, this work underlines some feedback on design and R&D aspects that should be addressed in the future to improve the SCT
Massimo Salvatores: integral experiments and their use for the validation of nuclear data and the neutronic design of advanced nuclear systems
No full textAmong the many domains of reactor physics on which Massimo Salvatores gave his considerable contributions, he was particularly passionate about integral experiments. In this paper, we make a review of selected experimental campaigns among the numerous ones he has promoted, conceived, designed, directed, or analyzed. They have been regrouped in a temporal sequence corresponding to the different periods of Massimo's career, which exceeded 50 years. When possible, for each of the experiments we provide a brief description, the goal for which it was conceived and carried out, and the practical impact on validation and design improvement. Finally, the conclusions offer thoughts and suggestions for the future of the integral experiments and a possible way of honoring the invaluable legacy that Massimo Salvatores has left to us