Summary
Optimization of nuclear fuel cycles is essential for experts and policy makers for studying and analyzing the future of the nuclear energy. In the case of advanced electronuclear transition ...scenarios, multiple parameters with a complex dependence have to be fine‐tuned in order to achieve a set of predefined objectives. However, in the presence of uncertainties the solutions obtained in this way may not be stable since small perturbations could break the delicate balance between different parts of the scenario. In this work, the optimization of an uncertainty European‐based sustainable transition scenario has been studied. This scenario, which has been analyzed with the TR_EVOL nuclear fuel cycle simulator system, is aimed at reducing the transuranic inventory masses while keeping the fuel cycle costs. To that end, an extension of the DEMO evolutionary multiobjective algorithm has been implemented within TR_EVOL for allowing the inclusion of constraints and uncertainties with a methodology that can be used by any fuel cycle simulator. Results show the importance of coupling optimization and uncertainty analyses due to the suboptimal and unstable solutions that can be obtained if not considered jointly. In addition, the uncertainties shrink the decision space. It was found that in their presence the transuranic mass can be reduced and stabilized by a factor ranging between 65% and 71% with an increase of the cost of 16% and 18.5% after 300 years of operation by using advanced systems when compared with an open fuel cycle strategy.
Novelty Statement
This work includes for the first time the coupling of multiobjective global optimization problems with uncertainties in the field of nuclear fuel cycle simulations. This problem, essential for strategy planning given the numerous assumptions and uncertainties surrounding nuclear sustainability studies, has proven to be crucial given that the optimal solutions cannot be reproduced in the presence of uncertainties. With the methodology presented, policy makers and experts can enhance confidence in their studies improving thus the sustainability of the nuclear energy.
The operation of accelerator-driven systems or spallation sources requires the monitoring of intense neutron fluxes, which may be billions-fold more intense than the fluxes obtained with usual ...radioactive sources. If a neutron detector is placed near a very intense source, it can become saturated because of detector dead time. On the contrary, if it is placed far away from the source, it will lose counting statistics. For this reason, there must exist an optimal position for placing the detector. The optimal position is defined as the one with the minimal relative uncertainty in the counting rate. In this work, we review the techniques to determine the detector dead time that can be applied with an accelerator-driven subcritical system or a spallation source. For the case of a spallation source, counting rates do not follow Poisson's statistics because of the multiplicity of the number of neutrons emitted by incident proton. It has been found a simple expression that relates the optimal counting rate with the source multiplicity and the uncertainty in the determination of the dead time.
The neutron-induced fission cross sections of Th-232 and U-233 were measured relative to U-235 in a wide neutron energy range up to 1 GeV (and from fission threshold in the case of Th-232, and from ...0.7 eV in case of U-233), using the white-spectrum neutron source at the CERN Neutron Time-of-Flight (n_TOF) facility. Parallel plate avalanche counters (PPACs) were used, installed at the Experimental Area 1 (EAR1), which is located at 185 m from the neutron spallation target. The anisotropic emission of fission fragments were taken into account in the detection efficiency by using, in the case of U-233, previous results available in EXFOR, whereas in the case of Th-232 these data were obtained from our measurement, using PPACs and targets tilted 45 degrees with respect to the neutron beam direction. Finally, the obtained results are compared with past measurements and major evaluated nuclear data libraries. Calculations using the high-energy reaction models INCL++ and ABLA07 were performed and some of their parameters were modified to reproduce the experimental results. At high energies, where no other neutron data exist, our results are compared with experimental data on proton-induced fission. Moreover, the dependence of the fission cross section at 1 GeV with the fissility parameter of the target nucleus is studied by combining those ( p, f) data with our (n, f) data on Th-232 and U-233 and on other isotopes studied earlier at n_TOF using the same experimental setup.
We report on the thermal neutron flux measurements carried out at the Laboratorio Subterráneo de Canfranc (LSC) with two commercial
2
″
×
2
″
CLYC detectors. The measurements were performed as part ...of an experimental campaign at LSC with
3
He detectors, for establishing the sensitivity limits and use of CLYCs in low background conditions. A careful characterization of the intrinsic
α
and
γ
-ray background in the detectors was required and done with dedicated measurements. It was found that the
α
activities in the two CLYC crystals differ by a factor of three, and the use of Monte Carlo simulations and a Bayesian unfolding method allowed us to determine the specific
α
activities from the
238
U and
232
Th decay chains. The simulations and unfolding also revealed that the
γ
-ray background registered in the detectors is dominated by the intrinsic activity of the components of the detector such as the aluminum housing and photo-multiplier and that the activity within the crystal is low in comparison. The data from the neutron flux measurements with the two detectors were analyzed with different methodologies: one based on an innovative
α
/neutron pulse shape discrimination method and one based on the subtraction of the intrinsic
α
background that masks the neutron signals in the region of interest. The neutron sensitivity of the CLYCs was calculated by Monte Carlo simulations with MCNP6 and GEANT4. The resulting thermal neutron fluxes are in good agreement with complementary flux measurement performed with
3
He detectors, but close to the detection limit imposed by the intrinsic
α
activity.
Celotno besedilo
Dostopno za:
DOBA, IZUM, KILJ, NUK, PILJ, PNG, SAZU, SIK, UILJ, UKNU, UL, UM, UPUK
The cross section of the
89
Y(n,
γ
) reaction has important implications in nuclear astrophysics and for advanced nuclear technology. Given its neutron magic number N = 50 and a consequent small ...neutron capture cross section,
89
Y represents one of the key nuclides for the stellar
s
-process. It acts as a bottleneck in the neutron capture chain between the Fe seed and the heavier elements. Moreover, it is located at the overlapping region, where both the weak and main
s
-process components take place.
89
Y, the only stable yttrium isotope, is also used in innovative nuclear reactors. Neutron capture and transmission measurements were performed at the time-of-flight facilities n_TOF at CERN and GELINA at JRC-Geel. Resonance parameters of individual resonances were extracted from a resonance analysis of the experimental transmission and capture yields, up to a neutron incident energy of 95 keV. Even though a comparison with results reported in the literature shows differences in resonance parameters, the present data are consistent with the Maxwellian averaged cross section suggested by the astrophysical database
KADoNiS
.
The Modified Source Multiplication method is used to determine an unknown reactivity level of a reactor from a known one if one has access to the detector counting for both levels when the reactor is ...fed by a constant neutron source like an Am-Be source. When available, an accelerator driven source, in continuous mode, can be useful as its intensity can be tunable and then adapted to the experimental conditions. However, in that case, the MSM technique must be extended to account for an external source whose intensity, energy and angular distributions can vary from one measurement to another. In this paper, this Modified Multi-Source Multiplication (MMSM) method is applied to measurements done during the FREYA project in the GUINEVERE facility, operated with the GENEPI-3C accelerator providing a mixture of (D,T) and (D,D) neutrons. The monitoring of these sources through the detection of the associated charged particles allows the calculation of the MMSM factors and the estimate of the reactivity values. The results are compared in different configurations with the reactivity obtained with an Am-Be source or in dynamic measurements performed with GENEPI-3C. Their excellent agreement shows the possibility of using such accelerator-based neutron sources for MSM measurements when they are correctly monitored. This is of great interest for deep sub-critical level characterization for which detector count rates per source neutrons might be low.
The neutron flux of the n_TOF facility at CERN was measured, after installation of the new spallation target, with four different systems based on three neutron-converting reactions, which represent ...accepted cross sections standards in different energy regions. A careful comparison and combination of the different measurements allowed us to reach an unprecedented accuracy on the energy dependence of the neutron flux in the very wide range (thermal to 1 GeV) that characterizes the n_TOF neutron beam. This is a pre-requisite for the high accuracy of cross section measurements at n_TOF. An unexpected anomaly in the neutron-induced fission cross section of
235
U is observed in the energy region between 10 and 30keV, hinting at a possible overestimation of this important cross section, well above currently assigned uncertainties.
The radiative capture cross section of a highly pure (99.999%), 6.125(2) grams and 9.56(5)E-4 atoms/barn areal density 238U sample has been measured with the Total Absorption Calorimeter (TAC) in the ...185 m flight path at the CERN neutron time-of-flight facility n_TOF. This measurement is in response to the NEA High Priority Request list, which demands an accuracy in this cross section of less than 3% below 25 keV. These data have undergone careful background subtraction, with special care being given to the background originating from neutrons scattered by the 238U sample. Pileup and dead-time effects have been corrected for. The measured cross section covers an energy range between 0.2 eV and 80 keV, with an accuracy that varies with neutron energy, being better than 4% below 25 keV and reaching at most 6% at higher energies.
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