The European Journal of Physics N (EPJ-N)
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Use of integral data assimilation and differential measurements as a contribution to improve
No full textCritical mass calculations of various HEU-fueled fast reactors result in large discrepancies in C/E values, depending on the nuclear data library used and the configuration modeled. Thus, it seems relevant to use integral experiments to try to reassess cross sections that might be responsible for such a dispersion in critical mass results. This work makes use of the Generalized Least Square method to solve Bayes equation, as implemented in the CONRAD code. Experimental database used includes ICSBEP Uranium based critical experiments and benefits from recent re-analyses of MASURCA and FCA-IX criticality experiments (with Monte-Carlo calculations) and of PROFIL irradiation experiments. These last ones provide very specific information on 235U and 238U capture cross sections. Due to high experimental uncertainties associated to fission spectra, we chose to consider either fitting these data or set them to JEFF-3.1.1 evaluations. The work focused on JEFF-3.1.1 235U and 238U evaluations and results presented in this paper for 235U capture and 238U capture, and inelastic cross sections are compared to recent differential experiment or recent evaluations. Our integral experiment assimilation work notably suggests a 30% decrease for 235U capture around 1–2.25 keV, a 10% increase in the unresolved resonance range when using JEFF-3.1.1 as “a priori” data. These results are in agreement with recent microscopic measurements from Danon et al. [Nucl. Sci. Eng. 187, 291 (2017)] and Jandel et al. [Phys. Rev. Lett. 109, 202506 (2012)]. For 238U cross sections, results are highly dependent on fission spectra
Innovative experiments for reduction of nuclear data uncertainty
No full textDue to the complexity of nuclear reaction models, current nuclear data evaluations must rely on experimental observations to constrain models and provide the accuracy needed for applications. For criticality applications, the accuracy of nuclear data needed is higher than what is currently possible from differential experiments alone, and integral measurements are often used for data adjustment within the uncertainties of differential experiments. This approach does not necessarily result in physically correct cross sections or other adjusted quantities because compensation between different materials is hard to avoid. One of the objectives of the recent CIELO project [M. Chadwick et al., Nucl. Data Sheets 118, 1 (2014)] was simultaneous evaluation of important materials in an attempt to minimize the effects of compensation. Improvement to the evaluation process depends on obtaining new experimental data with high accuracy and lower uncertainty that will help constrain the evaluations for certain important reactions. Improved experiments are accomplished by careful design with the objective of achieving high accuracy and lower uncertainty, and by designing new innovative experiments. New and unconventional experiments do not necessarily provide differential data but instead nuclear data that evaluators will find useful to constrain the evaluation and reduce the uncertainty. This also means that closer information exchange and collaboration between experimentalists and evaluators is important. For conventional experiments such as neutron transmission or capture measurements, it is important to understand the sources of uncertainty and address them in the experiment design. Such a process can also lead to the design of innovative methods. For example, the filtered beam method minimizes uncertainties due to background, and the Quasi-Differential Neutron Scattering method simplifies the experiment and data analysis and results in lower experimental uncertainty. A review of the sources of uncertainty in various experiments and examples of experimental techniques that help reduce experimental and evaluation uncertainty and increase accuracy will be discussed
Critical review of CIELO evaluations of
No full textKey reactions have been selected to compare JEFF-3.3 (CIELO 2) and IAEA CIELO (CIELO 1) evaluated nuclear data files for neutron induced reactions on 235U and 238U targets. IAEA CIELO evaluation uses reaction models to construct the evaluation prior, but strongly relied on differential data including all reaction cross sections fitted within the IAEA Neutron Standards project. The JEFF-3.3 evaluation relied on a mix of differential and integral data with strong contribution from nuclear reaction modelling. Differences in evaluations are discussed; a better reproduction of differential data for the IAEA CIELO evaluation is shown for key reaction channels
Production and use of nuclear parameter covariance data: an overview of challenging cross cutting scientific issues
No full textNuclear data users’ requirements for uncertainty data started already in the seventies, when several fast reactor projects did use extensively “statistical data adjustments” to meet data improvement for core and shielding design. However, it was only ∼20–30 years later that a major effort started to produce scientifically based covariance data and in particular since ∼2005. Most work has been done since then with spectacular achievements and enhanced understanding both of the uncertainty evaluation process and of the data utilization in V&V. This paper summarizes some key developments and still open challenges
Nuclear data adjustment based on the interpretation of post-irradiation experiments with the DARWIN2.3 package
No full textDARWIN2.3 is the French reference package dedicated to fuel cycle applications, computing fuel inventory as well as decay heat, neutron emissions, α, β and γ spectra. The DARWIN2.3 package fuel inventory calculation was experimentally validated with Post-Irradiation Experiments (PIEs), mainly consisting in irradiated fuel pellets analysis. This paper presents a method to assimilate these integral trends for improving nuclear data. In this study, the method is applied to 137Cs/238U concentration ratio. Results suggest an increase of the JEFF-3.1.1 235U cumulated thermal fission yield in 137Cs by (+3.8 ± 2.1)%, from 6.221E-02 to 6.460E-02 ± 2.1%
Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data
No full textThis paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal
On the influence of the americium isotopic vector on the cooling time of minor actinides bearing blankets in fast reactors
No full textIn the heterogeneous minor actinides transmutation approach, the nuclei to be transmuted are loaded in dedicated targets often located at the core periphery, so that long-lived heavy nuclides are turned into shorter-lived fission products by fission. To compensate for low flux level at the core periphery, the minor actinides content in the targets is set relatively high (around 20 at.%), which has a negative impact on the reprocessing of the targets due to their important decay heat level. After a complete analysis of the main contributors to the heat load of the irradiated targets, it is shown here that the choice of the reprocessing order of the various feeds of americium from the fuel cycle depends on the actual limit for fuel reprocessing. If reprocessing of hot targets is possible, it is more interesting to reprocess first the americium feed with a high 243Am content in order to limit the total cooling time of the targets, while if reprocessing of targets is limited by their decay heat, it is more interesting to wait for an increase in the 241Am content before loading the americium in the core. An optimization of the reprocessing order appears to lead to a decrease of the total cooling time by 15 years compared to a situation where all the americium feeds are mixed together when two feeds from SFR are considered with a high reprocessing limit
Uncertainty and covariances of the newly derived 8-groups delayed-neutrons abundances set
No full textDelayed-neutrons are of great importance for reactor operations. Current abundances derive from either a measurement performed in 1957 by Keepin or by a summation calculation performed by Brady and England in 1989. In this work, a code has been written to compute a new set of delayed-neutron abundances as well as to estimate uncertainties and correlations through a Monte Carlo method and a Bayesian inference. An experiment will take place in the future to verify the validity of the calculated quantities
Neutronics characterization of an erbia fully poisoned PWR assembly by means of the APOLLO2 code
No full textRecently, increasing demands on the reduction of fuel cycle costs have led to higher burnup fuel designs. According to the erbia-credit super high burnup fuel concept, developed by mixing low content of erbia to UO2 powder directly after reconversion process so that all fuel pins in a given fuel assembly are homogeneously doped, the present study aims to characterize, from a neutronic point of view, a 17 × 17 pressurized water reactor assembly enriched to 10.27 wt.% in 235U with an erbia content of 1 at.% (i.e. 0.7 wt.%) by means of the deterministic neutronic code APOLLO2. For this purpose, a simplified thermal-hydraulic analysis was performed in order to evaluate the effects on fuel thermal conductivity of adding erbia to uranium oxide. The results obtained allow to conclude that an Er-doped assembly enriched to >5 wt.% in 235U represents an advantageous solution for very long fuel cycles, and it is so suited for very high burnups
Proposal of new oxidation kinetics for sponge base E110 cladding tubes material
No full textStudy of high temperature steam oxidation kinetics during the high temperature oxidation was carried out on the sponge base E110 cladding tubes material in the temperature range 600–1300 °C. The oxidation kinetics derived from the weight gain measurements showed a parabolic rate law for temperatures 1100 °C and higher only. For lower temperatures in range 800–1050 °C especially, the parabolic law leads to very conservative prediction. Therefore, the new oxidation kinetics, different from the parabolic law, was designed. The experimental database containing more than 800 data points was compared with the new developed UJP-correlation and available correlations for E110 and Zircaloy alloys. Statistical analysis for all tested correlations was provided