A precise knowledge of the 235U prompt fission neutron spectrum is essential for reliable criticality safety analyses, reactor physics calculations, and nuclear data validation. In this work, the 235U prompt fission neutron spectrum was investigated and validated using a set of selected threshold reactions with well-known cross sections. Activation experiments were performed at the BR1 graphite-moderated research reactor using the MARK III 235U fission neutron converter and the S84 pneumatic irradiation channel. Threshold reactions 28Si(n,p)28Al, 169Tm(n,2n)168Tm, 92Mo(n,p)92mNb, and 19F(n,2n)18F were used for PFNS validation, 29Si(n,p)29Al, 141Pr(n,2n)140Pr, and 117Sn(n,n')117mSn reactions were selected for validation of their neutron cross sections, while capture reactions 176Yb(n,γ)177Yb, 141Pr(n,γ)142Pr, and 115In(n,γ)116mIn served for cross-section validation in thermal and epithermal neutron fields. Spectrum averaged cross section in 235U PFNS was measured for the first time for 141Pr(n,2n)140Pr reaction. Reaction rates were determined by high-resolution HPGe gamma-ray spectroscopy, applying efficiency transfer, coincidence summing, and self-shielding corrections supported by MCNP6.3 simulations. The measured spectrum-averaged cross sections were compared with calculations using IRDFF-II and ENDF/B-VIII.0 nuclear data libraries. Good agreement within combined uncertainties was observed for most dosimetric reactions, confirming the consistency of the ENDF/B-VIII.0 adopted 235U PFNS. However, significant discrepancies were identified for the silicon (n,p) reactions, indicating a potential need for reevaluation of these cross sections.

The dual-mode elpasolite scintillation material CLLBC (Cs2LiLaBr4.8Cl1.2:Ce) is capable of measuring both y-rays and neutrons. The neutron detection capability spans from thermal energies up to about 10 MeV, making these detectors attractive options for studying prompt fission neutron spectra (PFNS). In this work, a comprehensive characterization of CLLBC detectors is performed. Three CLLBC detectors were characterized, in addition to three LaBr3:Ce and two LaBr3:Ce,Sr for comparison. For the best-performing CLLBC detector, the results indicate an energy resolution of 3.7% at Ey = 662 keV and an intrinsic timing resolution of 1.2 ns (FWHM) above Ey = 1 MeV using 60Co. A y-neutron separation figure-of-merit of 2.7 is obtained by means of pulse-shape discrimination. Tagged neutron time-of-flight measurements were conducted using a 241Am9Be neutron source, by coincident detection of the 4.44 MeV y-ray and the neutron, to determine the intrinsic neutron detection efficiency between 2-6 MeV. Neutron detection efficiencies of about 0.2% for the 6Li(n,t)4He reaction and 1% for three types of (n,n ') reactions were obtained. Two of three investigated CLLBC detectors exhibit an energy peak asymmetry, resulting in worse performance, indicating scintillator quality issues and motivating further investigation. Future studies are anticipated using the 252Cf(sf) prompt fission neutron spectrum to determine neutron efficiencies for a wider range of neutron energies. Although observed in a previous study, neutron detection via 35Cl(n,p)35S was not identified in this work but is planned to be determined using quasi-monoenergetic neutrons generated at the JRC MONNET facility.

We present the findings of the probabilistic radiological characterization exercise conducted within the EU project MICADO ``Measurement and Instrumentation for Cleaning and Decommissioning Operations". A Bayesian inversion approach that accounts for uncertainty in the measurement efficiencies is used to interpret combinations of (i) segmented gamma scanning (SGS) spectrometry, (ii) passive neutron coincidence counting (PNCC) and (iii) active neutron interrogation (AN), in a fully virtual experiment. The considered Bayesian approach treats uncertainty in the measurement efficiencies by doing multilinear interpolation between reference efficiencies representing potential ``end-member" waste matrices with respect to both composition and density, with the end-member proportions being jointly inferred with the other unknowns. The performance of the approach in terms of efficiency and accuracy is explored for two virtual case studies of increasing complexity, that are based on common, real waste packages. The used Bayesian approach appears to be fast and rather accurate for the first considered waste package. With respect to the second waste package which has a substantially more complex and heterogeneous matrix structure, some biases are noticed in the derived posterior mass distributions of the nuclides of interest. We thus discuss possible causes and solutions for these discrepancies. In addition, we devise an R package that wraps the probabilistic models in an HTTP API so that the user can send HTTP requests to a remote server that runs the computations and returns the obtained results. This should allow using the approach in a production environment.

One of the main tasks in nuclear safeguards is regular inspections of Spent Nuclear Fuel (SNF) assemblies to detect possible diversions of special nuclear material such as 235U and 239Pu. In these inspections, characteristic signatures of SNF such as emissions of neutrons and gamma rays from the radioactive decay, are measured and their consistency with the declared assemblies is verified to ensure that no fuel pins have been removed. Research in this field is focused on both the development of detection equipment and methods for the analysis of the acquired measurement data. In this paper, the use of the neutron flux gradient, which is not considered in regular SNF verification, is investigated in combination with the scalar neutron flux as input to artificial neural network models for the quantification of fuel pins in SNF assemblies. The training and testing of these ANN models rely on a synthetic dataset that is generated from Monte Carlo simulations of a typical intact pressurized water reactor assembly and with different patterns of fuel pins replaced by dummy pins. The dataset consists of unique scenarios so that the ANN can be assessed over “unknown” cases that are not part of the learning phase. Results show that the neutron flux gradient is advantageous for a more accurate reconstruction of diversions within SNF assemblies.

The neutron-induced fission cross section of Th-230 has been measured at the neutron time-of-flight facility n_TOF located at CERN. The experiment was performed at the experimental area EAR-1 with a neutron flight path of 185 m, using Micromegas detectors for the detection of the fission fragments. The Th-230(n, f ) cross section was determined relative to the U-235(n, f ) one, covering the energy range from the fission threshold up to 400 MeV. The results from the present work are compared with existing cross-section datasets and the observed discrepancies are discussed and analyzed. Finally, using the code EMPIRE 3.2.3 a theoretical study, based on the statistical model, was performed leading to a satisfactory reproduction of the experimental results with the proper tuning of the respective parameters, while for incident neutron energy beyond 200 MeV the fission of( 230)Th was described by Monte Carlo simulations.

In this paper, we present the lessons learned from a multi-method, multi-mockup probabilistic radiological characterization exercise conducted within the EU project CHANCE "Characterization of Conditioned Nuclear Waste for its Safe Disposal in Europe". A Bayesian approach that accounts for uncertainty in the measurement efficiencies is used to interpret combinations of (i) open geometry gamma spectrometry (OGGS), (ii) passive neutron coincidence counting (PNCC) and (iii) large-volume calorimetry (LVC) measurements, performed on five different combinations of mockup drum and source position. The used approach treats uncertainty in the measurement efficiencies by doing multilinear interpolation between reference efficiencies representing potential "end-member" matrices with respect to both composition and density, for the considered drum. The end-member proportions are then jointly inferred with the other unknowns. Our results indicate that the success of this method critically depends on the accuracy of the set of reference end-member efficiencies obtained from physics-based Monte Carlo particle transport modeling. We find that Bayesian inversion of OGGS data alone can recover reasonably accurate nuclide masses, with relatively high precision and low bias. Furthermore, when little prior information about the drum is available then jointly inverting OGGS and PNCC data is found to allow for a much higher precision compared to inverting OGGS data alone. In addition, calorimetry appears to be a promising method and brings some added value to OGGS as well in certain cases. In some setups and for some nuclides, errors in the reference efficiencies cause the marginal posterior distribution to be fairly biased. We thus design a simple approach to increase posterior uncertainty (that is, decrease precision) such that the true mass values are included in the bulk of the posterior distribution, in a controlled way. The approach consists of multiplying the interpolated efficiencies with a series of multipliers, that are jointly inferred under a relatively strong prior.

During the course of the work carried out for the estimation of the sensitivity of Li-6 loaded fiber detectors (Borella, 2022), it was realized that the MCNP simulations carried out with the approach described in (Borella et al., 2013) were affected by a systematic bias. In this paper, we explain the source of the systematic bias and report the new results for (Borella et al., 2013).

We present three methodological improvements of our recently proposed approach for Bayesian inference of the radionuclide inventory in radioactive waste drums, from radiological measurements. First we resort to the Dirichlet distribution for the prior distribution of the isotopic vector. The Dirichlet distribution possesses the attractive property that the elements of its vector samples sum up to 1. Second, we demonstrate that such Dirichlet priors can be incorporated within an hierarchical modeling of the prior uncertainty in the isotopic vector, when prior information about isotopic composition is available. Our used Bayesian hierarchical modeling framework makes use of this available information but also acknowledges its uncertainty by letting to a controlled extent the information content of the indirect measurement data (i.e., gamma and neutron counts) shape the actual prior distribution of the isotopic vector. Third, we propose to regularize the Bayesian inversion by using Gaussian process (GP) prior modeling when inferring 1D spatially-distributed mass or, equivalently, activity distributions. As of uncertainty in the efficiencies, we keep using the same stylized drum modeling approach as proposed in our previous work to account for the source distribution uncertainty across the vertical direction of the drum. A series of synthetic tests followed by application to a real waste drum show that combining hierarchical modeling of the prior isotopic composition uncertainty together with GP prior modeling of the vertical Pu profile across the drum works well. We also find that our GP prior can handles both cases with and without spatial correlation. Of course, our GP prior modeling framework only makes sense in the context of spatial inference. Furthermore, the computational times involved by our approach are on the order of a few hours, say about 2, to provide uncertainty estimates for all variables of interest in the considered inverse problem. This warrants further investigations to speed up the inference.

In this paper neutron transport simulation results of experimental configurations used during the 2019 IPNDV measurement campaign at the Belgian Nuclear Research Centre SCK CEN in Mol are presented. As these are a good basis for benchmarking simulations, results of the Monte-Carlo simulation codes Geant4, SCALE/KENOVI, MCNP and openMC are compared with the aim of validating Geant4. With fissile material present as plutonium-uranium mixed oxide, (alpha, n) emissions are included in the simulations. A Geant4 extension to calculate neutron multiplication factors k(eff) was developed and applied to the IPNDV configurations and a set of OECD/NEA benchmark experiments. Neutron fluxes of the IPNDV configurations through a reference volume are simulated. The Geant4 (alpha,n) simulation toolkit SaG4n is compared to the SCALE code ORIGEN and small deviations are observed. Geant4 calculations of k(eff) for criticality benchmark experiments deviate in the mean by +0.4%, for MCNP and openMC deviations are <= 6%, for KENO-VI <= 8%. Geant4's total neutron fluxes through a reference volume next to the IPNDV configurations agree within a margin of +/- 5% with MCNP and openMC and -14 % with KENO-VI. Considerable differences between all codes were observed for thermal neutron scattering but Geant4 results are still within the other codes' variability. Differences to Geant4 for epithermal energies can be addressed to a not yet included ''Doppler Broadening Rejection Correction''. By comparing Geant4 simulations of a complex set-up with various reference codes and demonstrating its good performance as well as some discrepancies, this study contributes to the validation of Geant4 neutron physics in fissile material systems and for nuclear disarmament verification simulations.

Spent nuclear fuel represents the majority of materials placed under nuclear safeguards today and it requires to be inspected and verified regularly to promptly detect any illegal diversion. Research is ongoing both on the development of non-destructive assay instruments and methods for data analysis in order to enhance the verification accuracy and reduce the inspection time. In this paper, two models based on Artificial Neural Networks (ANNs) are studied to process measurements from the Partial Defect Tester (PDET) in spent fuel assemblies of Pressurized Water Reactors (PWRs), and thus to identify at different levels of detail whether nuclear fuel has been replaced with dummy pins or not. The first model provides an estimation of the percentage of replaced fuel pins within the inspected fuel assembly, while the second model determines the exact configuration of the replaced fuel pins. The two models are trained and tested using a dataset of Monte-Carlo simulated PDET responses for intact spent PWR fuel assemblies and a variety of hypothetical diversion scenarios. The first model classifies fuel assemblies according to the percentage of diverted fuel with a high accuracy (96.5%). The second model reconstructs the correct configuration for 57.5% of the fuel assemblies available in the dataset and still retrieves meaningful information of the diversion pattern in many of the misclassified cases.

A measurement campaign in support of nuclear disarmament verification was carried out in 2019 on the premises of the Belgian Nuclear Research Centre, SCK CEN. Well-characterized MOX sources with different types of shielding materials were used during the campaign. The relative content of Pu was up to 14%wt and the isotopic abundance of 239 Pu was up to 93%wt. One of the technologies that were deployed was neutron coincidence counting with 3 He based slab counters. The obtained background and dead time corrected total and reals rates were used to determine the mass of the assayed sample by using the so-called point model equations. The obtained masses were systematically underestimated by 15 % in average. A calibration approach based on Monte Carlo simulations where the mass is estimated from the calculated number of reals per spontaneous fission event was also used. A Monte Carlo model of the measurements setup was therefore developed to account for the finite source size, multiplication, moderation and absorption in the detector and scattering in the environment, including the measurement room walls, floor and ceiling. By using a Monte Carlo based efficiency for the reals the masses were slightly overestimated by typically few percent. However, outliers were observed. In an effort to understand the obtained results, additional calculations were carried out to estimate the impact due to both the neutron emission by (>,n) reactions and spontaneous fission decay on the measured observables. The associated total and reals counts were obtained for various configurations and compared with the experimental data. The obtained results indicate the importance of (>,n) reactions when neutron multiplication is not negligible.

•Systematic overestimation discovered in previously reported fission chamber neutron sensitivity data.•Tallying neutron flux from Maxwellian distribution source in MCNP was identified as the source of error.•New procedure to properly simulate neutron moderation was implemented.•Corrected results are reported.

A concept model of a measurement station for gamma-ray spectroscopy for safeguards verification of spent fuel from the future MYRRHA facility has been developed using MCNP6.2. The measurement station consists of a collimator and a lanthanum bromide detector which are placed adjacent to a cylindrical container with the fuel assembly submerged in lead–bismuth eutectic for cooling. The total count rate as well as the peak-area uncertainties for several gamma-ray peaks have been studied for a number of measurement geometries to determine the optimal collimator dimensions and assess how the performance depends on the lead–bismuth container geometry. Finally, the contribution from each fuel pin in the assembly to the detected signal has been determined as a function of gamma-ray energy to determine if the inner parts of the fuel assembly can be assayed with gamma-ray measurements. The results show that the dimensions of the lead–bismuth container will have a considerable effect on both the total count rate and the peak-area uncertainties, but that gamma-ray spectroscopy could be used for verification of spent fuel from MYRRHA.

Identification of the position of a localized neutron source, or that of local inhomogeneities in a multiplying or scattering medium (such as the presence of small, strong absorbers) is possible by measurement of the neutron flux in several spatial points, and applying an unfolding procedure. It was suggested earlier, and it was confirmed by both simulations and pilot measurements, that if, in addition to the usually measured scalar (angularly integrated) flux, the neutron current vector or its diffusion approximation (the flux gradient vector) is also considered, the efficiency and accuracy of the unfolding procedure is significantly enhanced. Therefore, in support of a recently started project, whose goal is to detect missing (replaced) fuel pins in a spent fuel assembly by non-intrusive methods, this idea is followed up. The development and use of a dedicated neutron detector for within-assembly measurements of the neutron scalar flux and its gradient are planned. The detector design is based on four small, fiber-mounted scintillation detector tips, arranged in a rectangular pattern. Such a detector is capable of measuring the two Cartesian components of the flux gradient vector in the horizontal plane. This paper presents an initial evaluation of the detector design, through Monte Carlo simulations in a hypothetical scenario.

Cadmium Zinc Telluride (CZT) detectors are portable, room temperature serviceable, medium-resolution gamma-ray spectrometers. Their full-energy peak shape exhibits a low energy tail which complicates the analysis of spectra with overlapping peaks. In this paper, we determined the peak shape parameters of a CZT detector from measurements with calibrated point sources from 0.06 MeV up to 1.332 MeV. The peak shape parameters were obtained by applying a peak fitting algorithm that includes a Gaussian and a tail with energy dependent parameters in the region around the gamma-ray peak. The net peak areas were used to verify the absolute detection efficiency obtained with a Monte Carlo model of the CZT detector and the agreement in absolute terms was within 10% over the considered energy range. The peak fitting algorithm was then applied to determine the net peak areas of the full energy peak in spectra recorded with certified uranium standards. The enrichment was then determined by using the so-called 'peak ratio' method. We observed a systematic bias in the net peak areas of the 0.258 MeV gamma-ray which therefore was not included in the analysis. Hence, the enrichment was underestimated by about 10%.

Within the EC-funded CHANCE project several non-destructive techniques are being considered for the assay of waste bearing drums. Such techniques include calorimetry, gamma-ray spectrometry and neutron coincidence counting. The aim is to quantify uncertainties on the inventory of radionuclides, and how these are potentially reduced by combining the signatures from different techniques in the data analysis. In this framework, neutron coincidence measurements were carried out with two slab counters based on 3 He detectors coupled to shift register electronics. Such a system consists of two identical slabs with 6 detectors each, and is transportable, rather compact and flexible in terms of sizes and geometries that can be measured. With this system three 200 L drums containing certified reference nuclear material and different filling materials were measured. The certified nuclear material was in the form of 21 pellets of mixed oxide of U and Pu with a total mass of about 10.5 g; in addition, a single pellet of about 10.05 g was also available. The pellets could be placed in predefined positions within the drum in a reproducible way. The geometry and composition of the three drums was well characterized and consisted of Ethafoam, a mixture of Ethafoam, stainless steel and PVC, and mortar with an inner core of extruded polystyrene. The measurement setup was arranged such that the drum was placed between the two slab counters. The positions of the slab counters relative to the drum were accurately measured before each measurement, and a dedicated system was used to minimize the uncertainty on the detector positioning. The measurement data were first analysed by applying the point model of Hage and the mass of nuclear material in the drum was determined from the rate of totals and reals and the radionuclide composition. Due to the fact that not all the point model conditions were met, we found that the point model overestimates the mass up to about 50%. In addition, a Monte Carlo model of the measurement geometry was developed using the MCNP code. The model was used to determine a calibration factor between the reals rate and the mass of the sample. Measurements with a calibrated 252 Cf source were used to verify the model. With a Monte Carlo based approach the mass of the mixed oxide pellets is within a few percent from the nominal values, except for strongly asymmetrical configurations where the deviation is up to about 20%. The results reveal the importance of an accurate background correction and of accounting for surrounding materials of the building such as walls, floor and ceiling in the Monte Carlo model.

Within the framework of safeguards verifications spent nuclear fuel is a concern because it contains nuclear material. Non-destructive assays (NDA) are amongst the safeguards measures for spent fuel verification. In this work machine learning using simulated data is investigated for the detection of fuel pin diversion. Three NDA techniques (Fork, SINRD, and PDET) and two machine learning approaches (decision trees and k-nearest neighbors) are considered to classify the assemblies according to the percentage of replaced pins. These NDA techniques combine different types of neutron and gamma-ray detectors. This study found that the classification accuracies using SINRD and PDET are higher compared to Fork. In addition, k-nearest neighbors models reached higher classification accuracies compared to decision tree models, and for the considered NDA techniques the gamma-ray detectors were the most sensitive to the fuel pin diversion. (C) 2020 Elsevier Ltd. All rights reserved.

A set of In-Pile Sections (IPS) has been irradiated in the BR2 reactor at SCK•CEN in Belgium during the 1970’s and 1980’s. The primary goal of the IPS was to replicate the thermo-hydraulic loop of a sodium-cooled fast reactor in order to study severe accident scenarios. The top part of the IPS contained the sodium-cooled loop whereas the lower part contained the fuel element. Due to the experimental conditions, the rupture of the fuel pins contained in the IPS occurred and fuel fragments may have been deposited in the rest of the IPS loop. The part of the IPS containing the fuel pins has been cut from the rest of the IPS and underwent post-irradiation examinations at specialized EU laboratories, while the top parts remained stored at SCK•CEN. To prepare for future transport, dismantling and conditioning, a reliable estimation of the total fissile content in the stored parts of the IPS is indispensable. In this framework, two IPS were measured with a Canberra WM3400 neutron coincidence counter with customized electronics. The measurements of the IPS were challenging due their length (roughly 6 m) and intense gamma-ray radiation background. For each IPS an axial scan was carried out with a series of short measurements (600-700 s each) recording the Totals rate and Reals rate. Based on the results of the axial scans, measurements with longer measurement time were conducted for the axial positions with the larger values of Reals rates. A system of equations was then established to quantify the 240Pu content in the different sections of the IPS from the Reals rates in each measurement position and account for cross-talk between the neutron emission associated to the different sections. A set of Monte Carlo simulations was carried out to estimate the probability to record a Real count in the detector due to spontaneous fission events occurring in a given section of the IPS. The 240Pu content in each section of the IPS was calculated by combining the measured Reals rates and the detection probabilities calculated with the simulations. The total fissile content in the IPS was then determined with scaling factors based on burnup calculations for the irradiated fuel assemblies in the IPS. The results indicate that both IPS measured with the neutron coincidence counters have a fissile content lower than the limit for transport. It is expected that the envisaged segmentation of the IPS in shorter sections required to fit into 200L drums will provide an additional safety margin on this limit.