Skin color-specific and spectrally-selective naked-eye dosimetry of UVA, B and C radiations

Spectrally–selective monitoring of ultraviolet radiations (UVR) is of paramount importance across diverse fields, including effective monitoring of excessive solar exposure. Current UV sensors cannot differentiate between UVA, B, and C, each of which has a remarkably different impact on human health. Here we show spectrally selective colorimetric monitoring of UVR by developing a photoelectrochromic ink that consists of a multi-redox polyoxometalate and an e− donor. We combine this ink with simple components such as filter paper and transparency sheets to fabricate low-cost sensors that provide naked-eye monitoring of UVR, even at low doses typically encountered during solar exposure. Importantly, the diverse UV tolerance of different skin colors demands personalized sensors. In this spirit, we demonstrate the customized design of robust real-time solar UV dosimeters to meet the specific need of different skin phototypes. These spectrally–selective UV sensors offer remarkable potential in managing the impact of UVR in our day-to-day life

The joint evaluated fission and fusion nuclear data library, JEFF-3.3

The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $^{235}$U, $^{238}$U and $^{239}$Pu, on $^{241}$Am and $^{23}$Na, $^{59}$Ni, Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy

recent progess in the vetv of the french apollo3 code 3d full core analysis of the uh1.2 integral experiment using idtmethod of characteristics

International audienceWithin the framework of the VetV of the new APOLLO3 French deterministic code, this paper focusses on its ability of analyzing integral mock-up experiments through IDT flux solver. In the present case, the UH1.2 experiments involving voiding effects in standard PWR UOX lattices are investigated through straightforward 281 groups 3D transport calculations using recent capabilities of IDT solver based on Sn short-characteristics method. Instead of using standard 2D calculations with imposed experimental et119861;et119911;2we performed directly 3D 281 groups IDT calculations and checked the results against reference continuous-energy Monte-Carlo (TRIPOLI4) results in a simplified axial configuration. In both cases the IDT solver (with dedicated self-shielding model) shows an excellent agreement with Monte-Carlo results regarding both reactivity and fission rate reference parameters. In particular it is shown that linear short characteristics and refined spatial meshes (3x3x1) are necessary to find the best agreement (less than 1percent fission rate and void effect errors). The next step of this work will be devoted to the analysis of the 3D TDT-MOC solver in the same configurations with ad-hoc axial reflector treatment. To remove the limitations in terms of CPU time and memory resources of this kind of calculations, ongoing developments devoted to hybrid shared/distributed parallelization (OpenMP/MPI) are also pointed out

Evaluation of Neutron-induced Cross Sections and their Related Covariances with Physical Constraints

International audienceNuclear data, along with numerical methods and the associated calculation schemes, continue to play a key role in reactor design, reactor core operating parameters calculations, fuel cycle management and criticality safety calculations. Due to the intensive use of Monte-Carlo calculations reducing numerical biases, the final accuracy of neutronic calculations increasingly depends on the quality of nuclear data used. This paper gives a broad picture of all ingredients treated by nuclear data evaluators during their analyses. After giving an introduction to nuclear data evaluation, we present implications of using the Bayesian inference to obtain evaluated cross sections and related uncertainties. In particular, a focus is made on systematic uncertainties appearing in the analysis of differential measurements as well as advantages and drawbacks one may encounter by analyzing integral experiments.The evaluation work is in general done independently in the resonance and in the continuum energy ranges giving rise to inconsistencies in evaluated files. For future evaluations on the whole energy range, we call attention to two innovative methods used to analyze several nuclear reaction models and impose constraints. Finally, we discuss suggestions for possible improvements in the evaluation process to master the quantification of uncertainties. These are associated with experiments (microscopic and integral), nuclear reaction theories and the Bayesian inference

From low- to high-energy nuclear data evaluations

Evaluation of neutron cross sections between 0eV and 20MeV is based on several aspects of nuclear physics such as nuclear reaction and structure models and microscopic and integral measurements. Most of the time, the evaluation process is separately done in the resolved resonance range and the continuum. It may give rise to non-physical mismatches of cross sections and large uncertainties at boundaries. It also leads to an absence of cross correlations between high-energy domain and resonance range. In addition, integral experiments are sometimes only used to check central values (evaluation is “working fine” on a dedicated set of benchmarks). Eventual reduction of uncertainties on cross sections is not straightforward: “working fine” could be mathematically turned into reduced uncertainties. This paper will present several ideas that could be used to avoid such effects. They are based on basic physical principles, recent advances in terms of covariance evaluation methodologies, intensive use of Monte Carlo methods and High Performance Computing (HPC) and on some newly introduced models. A clear connection is made between resonance and continuum energy ranges

Generation of

A new IAEA Coordinated Research Project (CRP) aims to test, validate and improve the IRDF library. Among the isotopes of interest, the modelisation of the 238U capture and fission cross sections represents a challenging task. A new description of the 238U neutrons induced reactions in the fast energy range is within progress in the frame of an IAEA evaluation consortium. The Nuclear Data group of Cadarache participates in this effort utilizing the 238U spectral indices measurements and Post Irradiated Experiments (PIE) carried out in the fast reactors MASURCA (CEA Cadarache) and PHENIX (CEA Marcoule). Such a collection of experimental results provides reliable integral information on the (n,γ) and (n,f) cross sections. This paper presents the Integral Data Assimilation (IDA) technique of the CONRAD code used to propagate the uncertainties of the integral data on the 238U cross sections of interest for dosimetry applications