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This explorative article aims to take a step in the direction of a realist-oriented scientific design that extends our knowledge of the requirements of a methodology that improves our ability to uncover the causal mechanisms behind men’s violence against women. Despite the great advances that have been made in individual research disciplines, our understanding of the complex causes is still insufficient and suffers from our inability to grasp the larger whole of the collaborative processes. As a first step towards the objective, an integration attempt is implemented that aims to highlight methodological issues that we have to overcome to explain men's violence against women. The integration of psychological, social-psychological, and sociological theories aims to exemplify how contributing, and counteracting factors interact with each other and form a complex mechanism that influences whether violence against women will take place or not. To leave room for the methodological dimension, the depth of each perspective has been reduced. The results of the integration attempt show both opportunities and difficulties in investigating the mechanisms behind men’s violence against women. However, there is still untapped knowledge potential in the explorative integration of theories and the use of realist-oriented pluralistic research methodologies.

Since Sweden has a reputation for having a pragmatic, technocratic approach to solving social problems, the question arises as to why the country uses different methods at the national and the local level for treating violent men. If studies show that one method is superior to others, we would expect both levels of government to use similar treatment methods. Despite the emphasis on pragmatic solutions, ideology plays an important role, as the Swedish government in recent decades has largely accepted the New Public Management approach to governance. However, because of differences in institutional arrangements, it becomes logically appropriate for the national level to utilize a different type of therapy (Cognitive Behavioral Therapy) than the local level (Psycho Dynamic Therapy among others). Thus, a combination of an ideological shift to New Public Management and institutional differences can explain the differences in therapeutic approaches, rather than reliance on scientific studies.

The aim of the present study is to deepen the understanding of over-indebted persons’ ways of describing the causes of their debts and the part they and others played in the emergence of the problematic situation, in their written communications with public authorities with influence over the debt-reconstruction process. The method was a qualitative multiple-case-study design with a purposive, cross-sectional sample consisting of personal letters attached to debt reconstruction applications submitted to the Swedish Enforcement Agency. The results reveal three main ways that applicants present their cases: Attributing responsibility to oneself, Avoiding the question of responsibility, and Attributing responsibility to others. The authors suggest that the results can be used in social work to help over-indebted people to understand and present their problems in a way that restores their vulnerable identity and allows them to counteract interpretations of them as having acted irresponsibly regarding their finances. Syftet med denna studie är att fördjupa förståelsen för överskuldsatta personers sätt att beskriva orsakerna till uppkomsten av deras skulder och till den roll de och andra spelade i uppkomsten av den problematiska situationen, i sin skriftliga kommunikation med myndigheter med inflytande över skuldsaneringsprocessen. Metoden genomfördes med hjälp av en kvalitativ multipel fallstudiedesign med ett målinriktat, tvärsnittsurval bestående av personliga brev bifogade skuldsaneringsansökningar till den svenska Kronofogdemyndigheten. Resultaten påvisar tre huvudsakliga sätt som de sökande presenterar sitt ärende på. I den första, Att tillskriva sig själv ansvar, hänvisar de sökande till sina egna handlingar i relation till skuldernas uppkomst på ett sätt som framställer dem som moraliskt värdiga personer. Det andra sättet att presentera sig själv är att Ge problemet egen agens. Det tredje och sista sättet är att Tillskriva andra ansvar.

We have proposed a planar electronically segmented HPGe detector concept in combination with a multi-slit collimator for gamma emission tomography. In this work, the spatial resolution achievable by using the collimated segmented HPGe detector was evaluated, prior to the manufacture and operation of the detector. The spatial response of a collimated segmented HPGe detector concept was evaluated using simulations performed with Monte Carlo N-Particle transport code MCNP6. The full detector and multi-slit collimator system were modeled and for the quantification of the spatial response, the Modulation Transfer Function (MTF) was chosen as a performance metric. The MTF curve was obtained through the calculation of the Line Spread Function (LSF) by analyzing simulated projection data. In addition, tomographic reconstructions of the simulated simplified test objects were made to demonstrate the performance of the segmented HPGe detector in the planned application. For 662 keV photons, the spatial resolution obtained was approximately the same as the collimator slit width for both 100 and 150 mm long collimators. The corresponding spatial resolution at 1596 keV photon energy was almost twice the slit width for 100 mm collimator, due to the partial penetration of the high-energy gamma rays through the collimator bulk. For a 150 mm long collimator, an improved resolution was obtained. VR Bränslediagnostik

A method to determine the absolute activity of 137Cs in irradiated nuclear fuel is presented. Using a well-known point-like calibration source in combination with measurements of the gamma-ray intensity from the nuclear fuel and Monte Carlo calculations based on the nominal measurement geometry, the activity content can be determined without prior knowledge of the intrinsic detection efficiency of the gamma-ray detector. The presented method is tested using measurements of the 137Cs intensity from spent nuclear fuel of the pressurized water type at the central interim storage in Sweden. Using an assumption of homogeneous distribution of 137Cs throughout the fuel, we demonstrate a linear relationship between measured activity and the activity calculated by a state-of-the-art simulation code. For future studies, we suggest some factors that potentially can decrease the uncertainty in the correlation between measured and calculated activity.

Glacier forelands contain valuable information on past glacier dynamics and associated climatic conditions, particularly at small mountain glaciers where responses to climate change are rapid. To maximize the potential of glacial landforms as palaeoclimate indicators, a thorough understanding of the controls on landform genesis and subsequent evolution is required. Traditionally, such landforms have been studied using glacial geological techniques such as sedimentary logging. While these provide valuable in situ information they have numerous limitations, namely poor availability and spatial extent of exposures. Near-surface geophysics provides an efficient and non-invasive means of studying subsurface conditions in numerous sedimentary settings, offering spatially extensive information on substrate material properties and architecture. However, the logistically challenging terrain, remote location and complex structure of proglacial environments has limited the development of geophysical techniques for studying the internal architecture of glacial landforms.Here, we explore the application of three geophysical methods to investigate proglacial substrates: ground penetrating radar (GPR), seismic refraction and multi-channel analysis of surface waves (MASW). Three sites with contrasting sediment properties were surveyed at the foreland of Midtdalsbreen glacier in southern Norway; (a) a 100 m2 area of glaciotectonised sandy sediments, (b) a ~2 m high lateral moraine ridge containing stratified silts, sands, and gravel and (c) a terminal moraine ridge with a peak crest height of ~5 m and an open blockwork of cobbles and boulders at its surface. At all sites, we deployed 25 MHz and 100 MHz GPR antennas and undertook seismic surveys with 50−75 m long geophone spreads and a sledge-hammer source to sample to target depths of around 10−15 m. Through comparing the results from sites (a) to (c), we assess the capabilities and limitations of each of the aforementioned techniques for proglacial substrate imaging and characterisation, we analyse how their performances vary across these settings and outline factors that contribute to a successful geophysical investigation. The ease of analysis and achievable investigation depths of the geophysical data and the applicability of seismic interpretation methods varied considerably depending on the surface terrain and structural complexity of the site. Our results show how the combination of GPR and seismic data can assist with the internal characterisation of glacial moraines when a relatively simple subsurface structure is present. However, basic seismic inversions likely lack the sophistication to resolve seismic structure in all but the simplest of layered models. We offer suggestions on how to optimise field time in more complex settings, where more sophisticated seismic inversion algorithms (e.g. tomography) or 3-D GPR surveys could be better-suited.Our experience should help advance the use of geophysics in proglacial studies. It should serve as a guide for future survey planning, and help avoid typical pitfalls such that field time can be optimised. It is hoped that geophysical survey methods will play an increasing role in the understanding of proglacial sedimentary landforms and their associated palaeoenvironments.

Gamma-ray spectrometry is widely applied in several science fields, and in particular in non-destructive gamma scanning and gamma emission tomography of irradiated nuclear fuel. Often, a collimator is used in the experimental setup, to selectively interrogate a region of interest in the fuel. For the optimization of instrument design, as well as for planning measurement campaigns, predictive models for the transmitted gamma-ray intensity through the collimator are needed. Commonly, Monte Carlo radiation transport tools are used for accurate prediction of gamma-ray transport, however, the long computation time requirements when used in low-efficiency experimental setups present challenges. In this work, the full-energy peak intensity transmitted through a rectangular collimator slit was examined. A uniform planar surface source emitting isotropically was considered, and the rate of photons reaching an ideal counter plane on the opposite side of the collimator was evaluated by analytical integration. To find a closed-form primitive function, some idealizations were required, and thereby parametric models were obtained for the optical field of view, dependent on slit dimensions (length, height and width) and source-to-collimator distance. It was shown that the count rate in the detector is independent of the collimator-to-source distance. For contributions from outside the optical field of view, where a closed-form expression cannot be found, instead fast numerical integration methods were proposed. The results were validated using the Monte Carlo code MCNP6.2. For the analytical method, deviations were larger, the shorter the collimator, with up to 25% of underestimation obtained for the shortest examined collimator of 10 cm length. However, the longer the collimator, the better the observed agreement. This accuracy is deemed to be sufficient for instrument design and measurement planning, where often the order of magnitude of the count rate is not a priori known. For the numerical method, the results showed an agreement within 3% for all evaluated collimator settings. The methods are planned for use in iterative optimization routines in the design of Gamma Emission Tomography devices, as well as for the prediction of gamma spectra obtained in the planning of fuel inspections. An application of the proposed method was demonstrated in spectrum prediction for a short cooling-time fuel rod test from the Halden reactor.

Segmented coaxial HPGe (High Purity Germanium) detectors have recently been shown to be feasible for Gamma Emission Tomography (GET). This type of detector allows for a combination of high efficiency and high energy resolution in gamma spectrometry of irradiated nuclear fuel. The ultimate aim of developing segmented HPGe for GET measurements is to achieve a high spatial resolution to facilitate imaging of rod-internal features and interrogation of smaller irradiated fuel samples. In this work, we present the optimisation of a segmented HPGe detector through a simulation study using the Monte Carlo particle transport code MCNP. Constraints to each dimension of the detector were identified, from manufacturing limitations and requirements arising from the use of a finite-sized collimator slit. In particular, a relationship between the minimum inner radius of the coaxial detector and the segments azimuthal dimension was derived based on the identified constraints. Segment azimuthal and radial dimensions have been varied (based on the derived relationship between the azimuthal and radial dimension) and the full energy efficiency and misidentification rate were evaluated to obtain the optimal dimensions. The optimal ranges of the segment dimensions were determined. VR Bränslediagnostik

Nondestructive gamma-ray spectrometry of nuclear fuel is routinely performed in axial gamma scanning devices and more recently with gamma emission tomography. Following the irradiation of a fresh nuclear fuel with high intensity neutron flux in a nuclear reactor core, a great number of gamma-emitting radionuclides are created. These can be utilized for gamma spectrometric techniques. However, due to the high density and atomic number of the nuclear fuel, self-attenuation of gamma-rays is a challenge, which requires attenuation correction in order to perform accurate analysis of the source activity in the fuel. In this study, the degradation of the gamma-ray mass attenuation with burnup was investigated and, in addition, a predictive model was created by investigating the attenuation change at various gamma energies caused by the burnup of the nuclear fuel. This model is intended for use by spectrometry practitioners inspecting nuclear fuel. To this aim, the energy-dependent gamma-ray mass-attenuation coefficients were investigated as a function of burnup for UOX, and three MOX fuels having different initial Pu contents. The Serpent 2 reactor physics code was used to simulate the burnup history of the fuel pins. The nuclide inventory of the Serpent 2 output is combined with the NIST XCOM database to calculate the mass attenuation coefficients. The mass attenuation coefficient of the fuel was found to decrease with the fuel burnup, in the range of a few percent, depending on the burnup and gamma energy. Also, a theoretical burnup dependent swelling model was imposed on fuel density to see how linear attenuation coefficient of fuel material is changed. Furthermore, greater effect may be expected on the transmitted intensity, where a simulation study of a PWR assembly revealed that the contribution from the inner rods in a scanned fuel assembly increased by tens of percent compared to the one with non-irradiated fresh fuels, when shielded by the outer rods of the assembly. A sensitivity analysis was performed in order to test the effect of a number of geometrical and operational reactor parameters that were considered to potentially effect the mass attenuation coefficient. Finally, a simple-to-use predictive model was constructed providing the mass-attenuation coefficient [cm2/g] of fuel as a function of burnup [MWd/kgHM] and initial Pu content [wt%]. The resulting predictive model was optimized by using the nonlinear regression method.

Irradiation testing of nuclear fuel is routinely performed in nuclear test reactors. For qualification and licensing of accident-tolerant fuels or generation IV reactor fuels, an extensive increase in irradiation testing is foreseen in order to fill the gaps of existing validation data, both in normal operational conditions and in order to identify operational limits. Gamma emission tomography (GET) has been demonstrated as a viable technique for studies of the behavior of irradiated nuclear fuel, e.g., measurement of fission gas release and inspection of fuel behavior under loss-of-coolant accident conditions. In this work, the aim is to improve the technique of GET for irradiated nuclear fuel, by developing a detector concept that allows for a higher spatial resolution and/or faster interrogation. We present the working principles of a novel concept for gamma emission tomography using a segmented high-purity germanium (HPGe) detector. The performance of this concept was investigated using the Monte Carlo particle transport code MCNP. In particular, the data analysis of the proposed detector was evaluated, and the performance, in terms of full energy efficiency and misidentification rate (i.e., localization failure), was assessed. We concluded that the segmented HPGe detector has an advantageous performance as compared to the traditional single-channel detector systems. Due to the scattering nature of gamma rays, a trade-off is presented between efficiency and cross-talk; however, the performance is nevertheless a substantial improvement over the currently used single-channel HPGe detector systems.

The Digital Cherenkov Viewing Device (DCVD) is a safeguards instrument available to international nuclear safeguards inspectors. It is frequently used to verify fuel on the gross defect level, and approved for partial defect verification, i.e. to assess that parts of a fuel assembly have not been diverted. The current limit for partial defect verification with the DCVD is on the 50% level. In the verification process, an analysis methodology is used where the inspector places a Region-Of-Interest (ROI) around the fuel assembly and assesses the total Cherenkov light intensity within this region. The intensity is then compared to a predicted value, and deviations from the predicted value are used to flag fuel assemblies for further investigations. In this work, we investigate a slightly different analysis approach, where the ROI is split into two or three segments to more accurately capture changes in light intensity in different regions of the captured image. The purpose is to increase the sensitivity of the DCVD to partial defects below the 50% level. Based on simulations of a Pressurised Water Reactor 17x17 fuel assembly, we conclude that a partial defect on the 30% level decreases the Cherenkov light intensity by at least 15% using one single ROI, by at least 20% using a ROI with two segments, and by at least 22% using a ROI with three segments. The analysis approach using two or three ROI segments instead of one thus appears to be more sensitive to partial defects, and can enable more accurate detection of partial defects on the 50% level as well as partial defect detection below the 50% level. Validation of the approach using a limited set of measurement data of intact fuel assemblies supports that detection of light intensity reductions by 20% and 22% is possible, while ensuring that the false positive rate is kept sufficiently low. However, an optimization of ROI segment splits as well as a more extended validation of the approach is required before the method can be considered reliable and applicable to all fuel assemblies that the DCVD can verify today.

Estimations of peatland carbon stocks often use generalised values for peat thickness and carbon content. Ground penetrating radar (GPR), a rapid technique for field data collection, has been increasingly demonstrated as an appropriate method of mapping peat thickness. Light Detection and Ranging (LiDAR) data as a method for understanding peatland surface elevation are also becoming more widely available. Reliable mapping and quantification of site-specific carbon stocks (e.g. upland raised bogs) is therefore, becoming increasingly feasible, providing a valuable contribution to regional, national and potentially global carbon stock assessments. This is particularly important because raised bogs, such as those found in South Wales are considerable carbon stores. They are, however, susceptible to climate warming owing to their southerly location within the UK. Accurate estimates of peatland carbon stocks has broader importance because world-wide peatland carbon stores are significant and threatened by climate change, posing a substantial challenge not only due to climate feedbacks if this stored carbon is released into the atmosphere, but also the impact on the other ecosystem services that they provide. Here, we assess the value of an integrated GPR, LiDAR and Geographic Information System (GIS) approach to improve estimation of regional carbon stocks. We apply the approach to three ombrotrophic raised bogs in South Wales, UK, selected for their conservation value and their topographically-confined raised bog form. GPR and LiDAR are found to be well suited, respectively, to mapping peat thickness at bog scale and surface elevation, thus allowing surface and basal topographies to be evaluated using GIS. In turn, this allows peat volumes to be estimated. For the first time, we record values between 55,200 m3 and 163,000 m3 for the sites considered here. The greater confidence in these peat volume estimates results from the ability to calibrate the GPR velocity using a depth-to-target calibration with peat cores extracted at locations encompassing the deepest bog area. Peat thickness is mapped at the bog scale with near centimetre precision, improving the robustness of subsequent volume calculations and our understanding of the contribution of these small but numerous sites to regional carbon stocks. Our evaluation shows that GPR corresponds well with conventional manual probing but is minimally invasive and therefore less disturbing of sensitive peatland sites, while also offering improved coverage and spatial resolution with less time and cost. In combination with measured bulk density and organic carbon contents, these peat volumes allow carbon stocks to be estimated with greater confidence compared to conventional approaches, having values between 2181 ± 122 tonnes carbon and 6305 ± 351 tonnes carbon at our three sites.

In this paper, we investigate the application of multivariate data analysis methods to the analysis of gamma spectroscopy measurements of spent nuclear fuel (SNF). Using a simulated irradiation and cooling of nuclear fuel over a wide range of cooling times (CT), total burnup at discharge (BU) and initial enrichments (IE) we investigate the possibilities of using a multivariate data analysis of the gamma ray emission signatures from the fuel to determine these fuel parameters. This is accomplished by training a multivariate analysis method on simulated data and then applying the method to simulated, but perturbed, data. We find that for SNF with CT less than about 20 years, a single gamma spectrum from a high resolution gamma spectrometer, such as a high-purity germanium spectrometer, allows for the determination of the above mentioned fuel parameters. Further, using measured gamma spectra from real SNF from Swedish pressurized light water reactors we were able to confirm the operator declared fuel parameters. In this case, a multivariate analysis trained on simulated data and applied to real data was used.

Raw data from calorimetric measurements of five nuclear fuel assemblies of the PWR 17 × 17 type are provided. Measurements of the temperature both inside a calorimeter, in which the fuel assembly was placed, as well as outside, were performed as a function of time while water circulating inside the calorimeter heats up from radiation emitted in the radioactive decay of material in the fuel assembly. The data contain also measurements of dose rate in the water outside the calorimeter. Data from 38 measurements using an electrically heated model of a fuel assembly are also provided to be used for, e.g., calibration. The data can be used for validation of computer codes for modelling of nuclear systems, e.g. nuclear reactors, storage and transport of nuclear fuel or systems for geological disposal.

The Digital Cherenkov Viewing Device (DCVD) is a tool used to verify irradiated nuclear fuel assemblies in wet storage by imaging the Cherenkov light produced by the radiation emitted from the assemblies. It is frequently used for partial defect verification, verifying that part of an assembly has not been removed and/or replaced. In one of the verification procedures used, the detected total Cherenkov light intensities from a set of assemblies are compared to predicted intensities, which are calculated using operator declarations for the assemblies. This work presents a new, time-efficient method to simulate DCVD images of fuel assemblies, allowing for estimations of the Cherenkov light production, transport and detection. Qualitatively, good agreement between simulated and measured images is demonstrated. Quantitatively, it is shown that relative intensity predictions based on simulated images are within 0.5% of corresponding predictions based solely on the production of Cherenkov light, neglecting light transport and detection. Consequently, in most cases it is sufficient to use predictions based on produced Cherenkov light, neglecting transport and detection, thus substantially reducing the time needed for simulations. In a verification campaign, assemblies are grouped according to their type, and the relative measured and predicted intensities are compared in a group. By determining transparency factors, describing the fraction of Cherenkov light that is blocked by the top plate of an assembly, it is possible to adjust predictions based on the production of Cherenkov light to take the effect of the top plate into account. This procedure allows assemblies of the same type bit with different top plates to be compared with increased accuracy. The effect of using predictions adjusted with transparency factors were assessed experimentally on a set of Pressurized Water Reactor 17x17 assemblies having five different top plate designs. As a result of the adjustment, the agreement between measured and predicted relative intensities for the whole data set was enhanced, resulting in a reduction of an RMSE from 14.1% to 10.7%. It is expected that further enhancements may be achieved by introducing more detailed top-plate and spacer descriptions.

Before encapsulation of spent nuclear fuel in a geological repository, the fuels need to be verified for safeguards purposes. This requirement applies to all spent fuel assemblies, including those with properties or designs that are especially challenging to verify. One such example are quivers, a new type of containers used to hold damaged spent fuel rods. After placing damaged rods inside the quivers, they are sealed with a thick lid and the water is removed. The lid is thick enough to significantly reduce the amount of the gamma radiation penetrating through it, which can make safeguards verification from the top using gamma techniques difficult. Considering that the number of quivers at storage facilities is foreseen to increase in near future, studying the feasibility of verification is timely. In this paper we make a feasibility study related to safeguards verification of quivers, aimed at investigating the gamma and neutron radiation field around a quiver designed by Westinghouse AB and filled with PWR fuel rods irradiated at the Swedish Ringhals site. A simplified geometry of the quiver and the detailed operational history of each rod are provided by Westinghouse and the reactor operator, respectively. The nuclide inventory of the rods placed in the quiver and the emission source terms are calculated with ORIGEN-ARP. The radiation transport is modeled with the Serpent2 Monte Carlo code. The first objective is to assess the capability of the spent fuel attribute tester (SFAT) to verify the content for nuclear safeguards purposes. The results show that the thick quiver lid attenuates the gamma radiation, thereby making gamma radiation based verification from above the quiver difficult.

In this paper we present a set of image analysis techniques used for extraction of information from cross-sectional images of nuclear fuel assemblies, achieved from gamma emission tomography measurements. These techniques are based on template matching, an established method for identifying objects with known properties in images. We demonstrate a rod template matching algorithm for identification and counting of the fuel rods present in the image. This technique may be applicable in nuclear safeguards inspections, because of the potential of verifying the presence of all fuel rods, or potentially discovering any that are missing. We also demonstrate the accurate determination of the position of a fuel assembly, or parts of the assembly, within the imaged area. Accurate knowledge of the assembly position enables detailed modelling of the gamma transport through the fuel, which in turn is needed to make tomographic reconstructions quantifying the activity in each fuel rod with high precision. Using the full gamma energy spectrum, details about the location of different gamma-emitting isotopes within the fuel assembly can be extracted. We also demonstrate the capability to determine the position of supporting parts of the nuclear fuel assembly through their attenuating effect on the gamma rays emitted from the fuel. Altogether this enhances the capabilities of non-destructive nuclear fuel characterization.

A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137 Cs, 154 Eu, and 134 Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden’s Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in themore » count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: 137Cs, 154Eu, 134Cs, and to a lesser extent, 106Ru and 144Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.« less

The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

As part of the United States (US) Department of Energy׳s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project, the traditional Differential Die-Away (DDA) method that was originally developed for waste drum assay has been investigated and modified to provide a novel application to characterize or verify spent nuclear fuel (SNF). Following the promising, yet largely theoretical and simulation based, research of physics aspects of the DDA technique applied to SNF assay during the early stages of the NGSI-SF project, the most recent effort has been focused on the practical aspects of developing the first fully functional and deployable DDA prototype instrument for spent fuel. As a result of the collaboration among US research institutions and Sweden, the opportunity to test the newly proposed instrument׳s performance with commercial grade SNF at the Swedish Interim Storage Facility (Clab) emerged. Therefore the design of this instrument prototype has to accommodate the requirements of the Swedish regulator as well as specific engineering constrains given by the unique industrial environment. Within this paper, we identify key components of the DDA based instrument and we present methodology for evaluation and the results of a selection of the most relevant design parameters in order to optimize the performance for a given application, i.e. test-deployment, including assay of 50 preselected spent nuclear fuel assemblies of both pressurized (PWR) as well as boiling (BWR) water reactor type.

In this study, snow particle size variability was investigated along a transect in Dronning Maud Land from the coast to the polar plateau. The aim of the study was to better understand the spatial and temporal variations in surface snow properties. Samples were collected twice daily during a traverse in 2007–08 to capture regional variability. Local variability was assessed by sampling in 10×10 m grids (5 m spacing) at selected locations. The particle size and shape distributions for each site were analysed through digital image analysis. Snow particle size variability is complex at different scales, and shows an internal variability of 0.18–3.31 mm depending on the sample type (surface, grid or pit). Relationships were verified between particle size and both elevation and distance to the coast (moisture source). Regional seasonal changes were also identified, particularly on the lower elevations of the polar plateau. This dataset may be used to quantitatively analyse the optical properties of surface snow for remote sensing. The details of the spatial and temporal variations observed in our data provide a basis for further studies of the complex and coupled processes affecting snow particle size and the interpretation of remote sensing of snow covered areas.

Using a high-purity Germanium gamma-ray energy spectroscopic detector system, time-stamped list-mode data sets were acquired during axial scanning of 19 boiling water reactor (BWR) and 28 pressurized water reactor (PWR) type of nuclear fuel assemblies. The data sets were collected during two measurements campaigns in September 2016 and March 2019 at the Central Interim Storage Facility for Spent Nuclear (Clab) in Sweden. A certified calibration source of 137Cs was positioned along the central line of sight between the measured fuel assembly and the detector. Data sets from measurements with only the calibration source and other background sources, i.e. without a nuclear fuel assembly present, are also included. The list-mode structure of the measured data allows for an axially-resolved as well as energy-spectral resolved intensity of nuclide-specific gamma lines emitted from the spent nuclear fuel. Data presented here can be used e.g. for validation of gamma-ray transport simulation tools or for development of methods to estimate parameters of the spent nuclear fuel based on data from gamma-ray spectroscopy.

The sixth edition of the International Conference on “Advancements in Nuclear Instrumentation Measurement Methods and their Applications” (ANIMMA) was held in Portorož, Slovenia, from June 17–21, 2019. The conference attracted almost 300 participants from 32 different countries, coming from academy, research institutes, and industry to discuss new scientific and technical prospects in all fields where nuclear instrumentation and measurements techniques play a major role. The scientific program included 25 invited talks, 125 oral presentations, and 77 poster presentations on the following topics: fundamental physics; fusion diagnostics and technology; nuclear power reactors monitoring and control; research reactors; nuclear fuel cycle; decommissioning, dismantling, and remote handling; safeguards and homeland security; severe accident monitoring; environmental and medical sciences; and education, training, and outreach.

Our understanding is limited to how transient changes in glacier response to climate warming will influence the catchment hydrology in the Arctic and Sub-Arctic. This understanding is particularly incomplete for flooding extremes because understanding the frequency of such unusual events requires long records of observation not often available for the Arctic and Sub-Arctic. This study presents a statistical analysis of trends in the magnitude and timing of flood extremes and the mean summer discharge in two sub-arctic catchments, Tarfala and Abisko, in northern Sweden. The catchments have different glacier covers (30% and 1%, respectively). Statistically significant trends (at the 5% level) were identified for both catchments on an annual and on a seasonal scale (3-months averages) using the Mann-Kendall trend test. Stationarity of flood records was tested by analyzing trends in the flood quantiles, using generalized least squares regression. Hydrologic trends were related to observed changes in the precipitation and air temperature, and were correlated with 3-months averaged climate pattern indices (e.g. North Atlantic oscillation). Both catchments showed a statistically significant increase in the annual mean air temperature over the comparison time period of 1985–2009 (Tarfala and Abisko pp = 0.91, Abisko p = 0.44). Despite the similar climate evolution over the studied period in the two catchments, data showed contrasting trends in the magnitude and timing of flood peaks and the mean summer discharge. Hydrologic trends indicated an amplification of the streamflow and flood response in the highly glacierized catchment and a dampening of the response in the non-glacierized catchment. The glacierized mountain catchment showed a statistically significant increasing trend in the flood magnitudes (p = 0.04) that is clearly correlated to the occurrence of extreme precipitation events. It also showed a significant increase in mean summer discharge (p = 0.0002), which is significantly correlated to the decrease in glacier mass balance and the increase in air temperature (p = 0.08). Conversely, the non-glacierized catchment showed a significant decrease in the mean summer discharge (p = 0.01), the flood magnitudes (p = 0.07) and an insignificant trend towards earlier flood occurrences (p = 0.53). These trends are explained by a reduction of the winter snow pack due to higher temperatures in the winter and spring and an increasing soil water storage capacity or catchment storage due to progressively thawing permafrost.

The Digital Cherenkov Viewing Device (DCVD) is an imaging tool used by authority inspectors for partial defect verification of nuclear fuel assemblies in wet storage, i.e. to verify that part of an assembly has not been diverted. One of the currently adopted verification procedures is based on quantitative measurements of the assembly's Cherenkov light emissions, and comparisons to an expected intensity, calculated based on operator declarations. A background subtraction of the intensity data in the recorded images is necessary for accurate quantitative measurements. The currently used background subtraction is aimed at removing an electronics-induced image-wide offset, but it is argued here that the currently adopted procedure may be insufficient. It is recommended that a standard dark-frame subtraction should be used, to remove systematic pixel-wise background due to the electronics, replacing the currently used offset procedure. Experimental analyses show that a dark-frame subtraction would further enhance the accuracy and reliability of DCVD measurements. Furthermore, should ageing of the CCD chip result in larger systematic pixel-wise deviations over time, a dark-frame subtraction can ensure reliable measurements regardless of the age of the CCD chip. It can also help in eliminating any adverse effects of malfunctioning pixels. In addition to the background from electronic noise, ways to compensate for background from neighbouring fuel assemblies and ambient light are also discussed.

Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242 Pu and 235 U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor. In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology.

Because climate change can affect the carbon balance and hydrology in permafrost peatlands, a better understanding of their sensitivity to changes in temperature and precipitation is needed. In Tavvavuoma, northernmost Sweden, meteorological parameters and ground thermal properties have been monitored in a peat plateau from 2006 to 2013. During this time period, the air temperature record shows no warming trend, and the late-season thaw depth has been relatively stable at around 55–60 cm. Meanwhile, the mean annual ground temperature at 1 m depth has increased by 0.06 °C/yr and at 2–5 m depth the permafrost is currently warmer than -0.3 °C. Statistical analyses suggest that interannual changes in thaw depth and ground temperatures are affected by different meteorological factors. Summer air temperatures and annual thawing degree-days control thaw depth (p ≤ 0.05), whereas winter precipitation/snow depth affects ground temperatures (p ≤ 0.1). The permafrost in this peat plateau is likely relict and not in equilibrium with current climatic conditions. Since the early 20th century, there has been a regional increase in air temperature and snow depth. If the ongoing permafrost warming in Tavvavuoma is a result of these long-term trends, short-term variability in meteorological parameters can still have an impact on the rate of permafrost degradation, but unless pronounced climate cooling occurs, thawing of the peat plateau is inevitable. Copyright © 2015 John Wiley & Sons, Ltd.

Previous simulation studies of Differential Die‐Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs. The results of this study suggest that DDA instrument response depends on the position of the individual neutron detectors and in fact can be split in two modes. The first mode, measured by the back detectors, is not significantly sensitive to the spatial distribution of fissile isotopes and neutron absorbers, but rather reflects the total amount of both contributors as in the cases of symmetrically burned SFAs. In contrary, the second mode, measured by the front detectors, yields certain sensitivity to the orientation of the asymmetrically burned SFA inside the assaying instrument. This study thus provides evidence that the DDA instrument can potentially be utilized as necessary in both ways, i.e. a quick determination of the average SFA characteristics in a single assay, as well as a more detailed characterization involving several DDA observables through assay of the SFA from all of its four sides that can possibly map the burn-up distribution and/or identify diversion or replacement of pins.

Over 11 000 L of kerosene was deposited on the surface of Rabots glaciär on the Kebnekaise Massif, northern Sweden, following the crash of a Royal Norwegian Air Force aircraft in March 2012. An environmental monitoring programme was subsequently commissioned, including a series of dye tracing experiments during the 2013 melt season, conducted to investigate the transport of pollutants through the glacier hydrological system. This experimental set-up provided a basis from which we could gain new insight into the internal hydrological system of Rabots glaciär. Results of dye tracing experiments reveal a degree of homogeneity in the topology of the drainage system throughout July and August, with an increase in efficiency as the season progresses, as reflected by decreasing temporary storage and dispersivity. Early onset of melting likely led to formation of an efficient, discrete drainage system early in the melt season, subject to decreasing sinuosity and braiding as the season progressed. Four distinct meltwater flow regimes are identified to summarize the temporal and spatial evolution of the system. Analysis of turbidity-discharge hysteresis further supports the formation of discrete, efficient drainage, with clockwise diurnal hysteresis suggesting easy mobilization of readily available sediments in channels. Dye injection immediately downstream of the pollution source zone reveals prolonged storage of dye followed by fast, efficient release. Twinned with a low dye recovery, and supported by sporadic detection of hydrocarbons in the proglacial river, we suggest that meltwater, and thus pollutants in solution, may be released periodically through an efficient, and likely pressurized, hydrological system within the upper reaches of the glacier.