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A technique for the in-situ localization of radioactivity is described in which the influence of γ-radiation impinging on a high-Z collimator, by which the angular response of a scintillation detector is constrained in order to identify the corresponding angular position of the radiation source, is expressed mathematically by way of a normalized sinc transform. We test this approach by examining the utility of the sinc transform to express the angular responses derived from a slot-collimated cerium bromide detector, across a variety of energy regions. Individual spectra have been acquired as a function of angle to explore how the shape of the response of the collimator-detector arrangement changes for X-rays and γ rays. A 90% improvement in localization is observed when this is defined in terms of the area described by the variance between the known location and that indicated by the response of the collimated system. This approach has the potential to improve source location accuracy and to further optimize autonomous robot exploration routines used to characterize contaminated environments associated with nuclear legacies and radiological emergencies.

Collection and interpolation of radiation observations is of vital importance to support routine operations in the nuclear sector globally, as well as for completing surveys during crisis response. To reduce exposure to ionizing radiation that human workers can be subjected to during such surveys, there is a strong desire to utilise robotic systems. Previous approaches to interpolate measurements taken from nuclear facilities to reconstruct radiological maps of an environment cannot be applied accurately to data collected from a robotic survey as they are unable to cope well with irregularly spaced, noisy, low count data. In this work, a novel approach to interpolating radiation measurements collected from a robot is proposed that overcomes the problems associated with sparse and noisy measurements. The proposed method integrates an appropriate kernel, benchmarked against the radiation transport code MCNP6, into the Gaussian Process Regression technique. The suitability of the proposed technique is demonstrated through its application to data collected from a bespoke robotic system used to conduct a survey of the Joz̆ef Stefan Institute TRIGA Mark II nuclear reactor during steady state operation, where it is shown to successfully reconstruct gamma dosimetry estimates in the reactor hall and aid in identifying sources of ionizing radiation.

We propose a method to predict the dose rate distribution in a primary containment vessel of Unit 1 of the Fukushima Daiichi Nuclear Power Station (1F) using numerical calculations and measured data.

This research presents a design of a resilient blind-tube radiometric logging probe, which satisfies key, operational site constraints, for direct in-situ characterization of subsurface radioactivity, particularly caesium-137 and strontium-90. The probe comprises a commercially-available Ø10 mm × 9.5 mm CeBr3 scintillator detector attached to a compact digitizer unit, in a resistant and waterproof housing. The probe is designed to be lowered down into metallic Ø 75 mm blind-tubes by means of a winch system.

Environmental radioactivity has been reported in the stationary wetlands closer to the natural water basins in Great Britain, precisely in the English Lake District (Cumbria, UK) often emphasise on the investigation carried out through either high-resolution γ-ray spectroscopy or α-particle spectroscopy. The objective of the study is the evaluation of trace actinide (241Am) through environments radioactivity measurements of two Lakes in the Lake District using a broad-energy, high-resolution γ-ray spectroscopy germanium (BEGe) detector. These appeal to the need to determine the radioactive concentration (Bq/g) in the soil samples with the intent being to discern natural and anthropogenic contributions, and trends in abundances associated with influences of the landscape over time, at trace levels in the environment. These measurements highlight potential benefits for the assessment of 241Am and 241Pu, especially comparisons that might be made with accelerator mass spectrometry assessments (AMS). This study intents to offer an inclusive investigation practise for determining the levels of radioactivity in soil cores, as well as the method for statistical analysis approach to calculate the activity of a soil sample, with a special emphasis on categorising the natural and anthropogenic contributors to trace amounts of plutonium in the environment, that is, less than500 fg/g.

Trace-level plutonium in the environment often comprises local and global contributions, and is usually anthropogenic in origin. Here, we report estimates of local and global contributions to trace-level plutonium in soil from a former, fast-breeder reactor site. The measured 240Pu/239Pu ratio is anomalously low, as per the reduced 240Pu yield expected in plutonium bred with fast neutrons. Anomalies in plutonium concentration and isotopic ratio suggest forensic insight into specific activities on site, such as clean-up or structural change. Local and global 239Pu contributions on-site are estimated at (34 ± 1)% and (66 ± 3)%, respectively, with mass concentrations of (183 ± 6) fg g−1 and (362 ± 13) fg g−1. The latter is consistent with levels at undisturbed and distant sites, (384 ± 44) fg g−1, where no local contribution is expected. The 240Pu/239Pu ratio for site-derived material is estimated at 0.05 ± 0.04. Our study demonstrates the multi-faceted potential of trace plutonium assay to inform clean-up strategies of fast breeder legacies., Nature Communications, 12 (1), ISSN:2041-1723

The design and test of a robot-compatible, radiation detection instrument providing simultaneous $\gamma $ -ray imaging and $\gamma $ -ray spectroscopy is described. The sensing system comprises a cerium bromide inorganic scintillation detector and a cylindrical, lead slot collimator that is configured with a robot-compatible, on-board data acquisition system. The mount for the sensor is a lightweight, bespoke 3-axis gimbal actuated by servos for pan, tilt and rotation of the collimator for imaging capability. This paper discusses the integration of this relatively low-cost radiation detection apparatus with a commercially available, Robot-Operating-System-controlled robotic platform (a Clearpath Robotics™ Jackal). The detection system is compliant with the power and mass payload constraints of the robot. Its performance has been evaluated by means of two practical examples: a) measurements in a laboratory environment to assess the ability of the system to resolve two caesium-137 point sources, and b) deployment at the Jožef Stefan Institute TRIGA Mark II research reactor to assess the ability of the system to characterise the $\gamma $ -ray emission at 1 kW from a horizontal tangential beam port in the reactor hall and from the reactor sample pool above the core.

The identification of corrosion, cracks and defects in pipelines used for transporting oil and gas can reduce the possibility of leaks, and consequently, it can limit the extent of an environmental disaster, public hazard and the associated financial impact of such events. Typically, corrosion in oil pipelines is measured with non-destructive ultrasonic or electromagnetic techniques, on the basis that corrosion and defects are often manifest as a change of thickness in the steel from which pipelines are made. However, such approaches are not practical for underground pipelines and their deployment can be complicated for the case of pipelines covered by insulation. In this paper, we present an innovative, non-destructive testing technique, which exploits the backscatter of a combination of fast-neutron and γ radiation from steel samples of a variety of thicknesses consistent with changes that might arise due to corrosion of a pipe wall. Our research demonstrates the potential to measure and characterise different steel thicknesses by detecting both the elastic, fast-neutron backscatter and the Compton-scattered γ radiations, simultaneously. Further, we demonstrate that the presence of insulation yields a consistent and separable influence on the experimental, wall-thickness measurements. The data from experimental measurements are supported by a comprehensive Monte Carlo computer simulation study.

High-order angular correlations of fast neutrons emitted by spontaneous fission in californium-252 ( 252 Cf) have been measured with a 2D array of 15 EJ-309 organic liquid scintillation detectors. These are compared with the results of Geant4 simulations using both an uncorrelated, average fission model and correlated fission models, which have also been used to quantify the impact of detector cross-talk on these distributions. In general, a bipolar trend is observed up to and including quadruple events, that is, θ 12 , θ 13 and θ 14 . A bias as θ → 0 ∘ in these data is observed that is consistent with the expectation of cross-talk being more prevalent for nearest-neighbour detectors. This is observed to be reduced for higher orders ( θ 13 and θ 14 relative to θ 12 ) and for wider gate widths (25 ns rather than 10 ns), consistent with this phenomenon being cross-talk. For the case of θ 13 , each of the composite angular distributions for the 14 detector positions 2 4 ∘ ≤ θ ≤ 33 6 ∘ have also been derived: a variation from a bipolar trend to a more isotropic behaviour, accompanied by a reduction in the relative level of response, is observed in these data for detector angles orthogonal to the reference detector at 0 ∘ consistent with the majority of neutrons being emitted from accelerated fragments. This research constitutes an important step towards quantifying and managing cross-talk as arrays of organic scintillators attract interest for practical applications such as nuclear materials assay.

In this paper, the irradiation of glycerol and ethylene glycol by either mixed (neutron-γ) or γ-only (γ) fields at the TRIGA reactor of the Jožef Stefan Institute is described. This is highly relevant to future applications of fission reactor systems to produce useful feedstock derivatives from organic waste, beyond the production of heat and power. Samples of glycerol and ethyl glycol have been exposed to neutron-gamma radiation with fast neutron fluxes ranging from 7.7×1010 to 3×1012 cm−2s−1 and gamma-only irradiation at maximum dose-rates of 492 and 10 kGy hr−1, respectively. A study of the dependence of product yield versus absorbed dose has been conducted, for total dose ranges of 1 to 100 kGy. The products of these exposures have been identified through Gas Chromatography – Mass Spectrometry (GC-MS) techniques. Analysis comparing neutron-gamma irradiated samples of ethylene glycol and glycerol with gamma-irradiated samples shows no detectable qualitative difference between either irradiation type. Although, additional radiolysis products were detected when compared with available literature; ethyl acetate from ethylene glycol and solketal from glycerol. Quantitatively, neutron-gamma irradiation seems to be less effective at producing acetaldehyde from ethylene glycol, compared with gamma-only which can be explained through neutron moderation and consequent heating effects due to the borosilicate vials.

Charge coupled devices (CCDs) have been shown to have potential for detecting charged particles and other forms of ionising radiation. In particular, the clusters in the pixel images produced are distinctive for α and β radiations, with α particles causing large, symmetrical clusters or long, vertical tracks often referred to as blooming, and β particles causing long, curved tracks. The size and shape of these tracks are also related to the energy of the incident radiation, giving the potential for spectroscopy of these types of radiation. This could be used, for example, to realise a hand-held, portable device for in-situ detection and identification of radioactive contamination. Images have been taken of exposures to a 210Po α-particle source using Sony ICX825AL interline transfer CCDs, covered with a thin, aluminised Mylar film to prevent inadvertent exposure to light. Both vertical streaks and round clusters have been observed in each of the three CCDs used. Increased levels of noise have been seen after several hours of exposure to the 210Po source, found to be caused by damage from α radiation. Three methods have been investigated to reduce this noise: 1) Annealing at 100°C for 24 hours repairs a significant proportion of the damage caused by the α radiation, though the effectiveness is reduced with each subsequent exposure and annealing cycle; 2) Cooling is performed during exposures by using a heat pipe to move heat away from the back of the CCD and a fan to cool the heat pipe and electronics associated with the CCD.This reduces the dark current during CCDmeasurements, and the method of cooling used allows thedevice to retain its small, hand-held size; 3) Anyremaining noise can be removed later using imageprocessing. These techniques are combined to extend thelifetime of the device. The rate of damage from the 210Posource is similar whether the device is cooled or not,however the noise evident in the images produced whencooled is significantly less, allowing the CCD to be usedfor a longer time between annealing operations.

Fast neutrons propagate significantly in free space with interaction properties that contrast markedly with those of electromagnetic radiation , the latter being fundamental to most wireless communications . Here, we report on the configuration and operation of a nuclear instrumentation set-up designed to transmit digitally encoded information using fast neutrons. We demonstrate the potential of fast neutron radiation as a medium for wireless communications for applications where conventional electromagnetic transmission is either not feasible or is inherently limited.

The use of robotics in harsh environments, such as nuclear decommissioning, has increased in recent years. Environments such as the Fukushima Daiichi accident site from 2011 and the Sellafield legacy ponds highlight the need for robotic systems capable of deployment in hazardous environments unsafe for human workers. To characterise these environments, it is important to develop robust and accurate localization systems that can be combined with mapping techniques to create 3D reconstructions of the unknown environment. This paper describes the development and experimental verification of a localization system for an underwater robot, which enabled the collection of sonar data to create 3D images of submerged simulated fuel debris. The system was demonstrated at the Naraha test facility, Fukushima prefecture, Japan. Using a camera with a bird&rsquo, s-eye view of the simulated primary containment vessel, the 3D position and attitude of the robot was obtained using coloured LED markers (active markers) on the robot, landmarks on the test-rig (passive markers), and a depth sensor on the robot. The successful reconstruction of a 3D image has been created through use of a robot operating system (ROS) node in real-time.

The use of ground-based robotic systems for the characterisation of nuclear environments is reviewed. Almost since the dawn of the nuclear energy industry, man has somewhat inadvertently created environments in which access has been constrained primarily due to the risk posed by extreme levels of radiation exposure but also due to space constraints, and because of toxic and combustible atmospheres. Robotic systems pose an ideal solution to some of these difficulties, removing the need for humans to access such places and frequently providing data on the state of such places that would not otherwise be available. However, each of these requirements is often very different in terms of the specification of a given robot, and the detailed characteristics of a given harsh environment can pose significant challenges even for the most robust of platforms. Furthermore, such developments can be expensive in terms of cost and development time. These issues notwithstanding, robotic solutions to nuclear challenges are reaching a level of maturity where their use is destined to add significant value. This paper considers the salient developments in ground-based solutions from the era preceding the Three Mile Island accident, through Chernobyl and on to the present day and, in particular, the needs of Fukushima Daiichi as attentions turn to this complex robotic suite of challenges.

The ability to detect neutrons from the spontaneous fission of 238U in samples of depleted uranium with organic liquid scintillation detectors is presented. In this paper we introduce a small modular organic liquid scintillator detector array that can detect changes in mass of 238U between 3.69 g and 14.46 g. To do this, 18-h assays of various masses of 0.3% wt. of depleted uranium dioxide were assessed using four EJ-309 detectors, a mixed field analyser operated in pulse gradient analysis mode, and associated counting components. We observe a background-corrected fast neutron count sensitivity of (2.0 ± 0.3) × 10−4 n g−1 s−1 per detector. This research demonstrates a proof of concept for depleted uranium quantity to be assessed passively on a non-intrusive basis via its spontaneous fission decay.

Fast neutron time-of-flight (ToF) has been used to characterise the location of a source of a mixed radiation field. Two EJ-309 organic scintillators and a fast, digital, data acquisition system have been used in a variety of positions to identify the location of a 252Cf neutron source inside a steel, water-filled tank. A methodology for extracting the distance between the neutron source and the neutron detector has been developed and verified with MCNP simulations. A reconstruction algorithm using the ToF data has been developed. The location of the neutron source has been estimated on this basis to be within 20 cm of its known location with a spatial resolution of ± 7.8 cm. p-values extracted from the null test hypothesis have been estimated to be 0.975 and 0.996 for experimental and simulation data, respectively. By correctly identifying the location of the source, the potential for the system to discern between scattered and unscattered neutrons is demonstrated.

The application and feasibility of a time-of-flight neutron detection system is explored for sources with time correlated gamma-ray and neutron emissions, such as the spontaneous fission emitter, Cf-252. For the emission of multiple gamma rays and neutrons from a single spontaneous fission event, a near instantaneous gamma-ray detection followed by a later neutron detection on a multi-detector array allows for an associated time-of-flight to be determined for a neutron arising from that event. Using a suite of purpose developed analysis tools, Monte-Carlo simulation and experimental data are compared for the Cf-252 water tank source facility at Lancaster University. Applying a bespoke time-of-flight imaging algorithm, vector-based optimisation (VBO), the true source location is determined within 21 cm by this approach.

The sensitivity of an active fast neutron assay system for the measurement of 235U enrichment in small samples of low-enriched triuranium octoxide (U3O8) is described. The system comprises four organic liquid scintillation detectors in a polyethylene container in which a stimulating americium-lithium neutron source is placed below the sample to be interrogated. The sensitivities of both the singles and doubles assay have been corrected for the contribution from the stimulating source and for spontaneous fission in 238U and found to be (0.116 ± 0.008) cps per detector per % wt. (7%) and (0.0006 ± 0.0002) cps per detector per % wt. (33%), respectively. The singles approach with the 4-detector arrangement has been compared via Monte Carlo (MCNP-5) simulations with a fast neutron counting system based on 12 detectors in which the source is placed adjacent to the sample, similar to the arrangement referred to as the liquid scintillator uranium neutron collar. This comparison confirms that whilst singles assay with the 4-detector arrangement is feasible, the collar arrangement does not yield a singles yield that can be correlated with enrichment due to the perturbative scattering by the U3O8 sample. The 4-detector arrangement is particularly suitable for small samples of fresh, low-enriched material that might arise in forensic applications and the analysis of un-irradiated, orphan wastes. In these cases the efficiency of coincidence methods may be too low to yield a practical route to enrichment assessment. Conversely, the use of many detectors and/or a high-intensity source by which sensitivity might be increased might restrict the portability and ease of use of the apparatus. A correction for the contribution by (α, n) reactions, the stimulating source, spontaneous fission in 238U and self-multiplication in U3O8 is made via preparatory passive measurements.

Shutdown dose rate calculations have been performed on an integrated ITER C-lite neutronics model with equatorial port 10. A ‘fully shielded’ configuration, optimised for a given set of diagnostic designs (i.e. shielding in all available space within the port plug drawers), results in a shutdown dose rate in the port interspace, from the activation of materials comprising equatorial port 10, in excess of 2000 μSv/h. Achieving dose rates of 100 μSv/h or less, as required in areas where hands-on maintenance can be performed, in the port interspace region will be challenging. A combination of methods will need to be implemented, such as reducing mass and/or the use of reduced activation steel in the port interspace, optimisation of the diagnostic designs and shielding of the port interspace floor. Further analysis is required to test these options and the ongoing design optimisation of the EP10 diagnostic systems.

The actinides are thought to have been generated by the rapid neutron capture (r)-process in a supernova at the origin of the solar system (4.6 × 109 an ago) or even before. Thus, given the generation processes for thorium, Th, and uranium, U, (and their precursors) are somewhat similar, the difference in half-lives (232Th: 14.05 × 109 a, and 238U: 4.468 × 109 a) would suggest a larger abundance of Th than U be present today. However, the chemistry of U is very different to that of Th which could suggest different pathways to explain the different concentrations of Th and U in various formations and local deposits of UO2. Focussing on the accessible geosphere of the Earth (i.e., the crust and hydrosphere), the evidence for the presence of these actinide elements in the liquid and solid rocks was examined. This study discusses why, in seawater, soluble Th (5 × 10−11 M) is more depleted than U (1.4 × 10−8 M), and why solid ores of pure UO2 or U3O8 are relatively common while ThO2 deposits are quasi-non-existent. Thorium is rather diluted in rock, relative to uranium, requiring that more mining, milling and treatment is necessary to recover it on comparable scales. This paper discusses and compares the abundances of U and Th in the universe and more specifically in the Earth's accessible geosphere as well as the extent of their known, economically-recoverable resources.

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.

Most techniques that are used for transmission imaging with ionizing radiation use X-rays, which have the advantage of providing quick, high-resolution images with a relatively small dose of radiation. However, they also have the disadvantage that their penetrating power can be limited in some forms of matter. This can make the discrimination of materials with a low atomic number particularly challenging. Of specific interest in this regard is the need to screen a diversity of manmade items that are heterogeneous and with the tendency to have many interfaces between components that can comprise a diversity of low-mass elements and compounds. These items usually have a compact geometry and a high density of components, which can make them less easy to be imaged quickly and effectively with X-rays. This limit of current screening technology necessitates further stages of examination reducing the ease with which this is done for manufacturing and quality assurance applications. The results presented in this paper demonstrate that, either via fast-neutron radiography or tomography, the potential exists to discern a variety of low-A compounds from one another. Via Monte Carlo simulations, it will be shown that fast-neutron radiography undertaken with a portable, isotopic radiation source (californium-252), absorption and scattering by the doped polymeric materials yields a degree of distinction from other substances. Considering these results, the state-of-the-art of the technique leading to the realization of a combined, real-time fast-neutron and gamma-ray radiography system will also be presented, as well as the first experimental results.

In this research, a combined fast-neutron/γ-raybackscatter imaging technique is described. The aim of this work is to understand corrosion defects in pipelines by measuring differences in the scattered radiation flux, generated when different steel thicknesses are irradiated by a neutron and γ-ray focused beam. A californium-252 radiation source is used to produce fast neutron and γ rays, exploiting its spontaneousfission. This mixed radiation field is collimated and directed towards the steel samples. Backscattered neutrons and γ rays aremeasured as a function of the steel thickness using 4 liquid organic scintillation detectors linked to a real-time, pulse-shape discrimination system, which separates and retains the neutron and γ-ray event data. In this paper, we describe how, using asingle radiation source and detection system, it is possible to perform and combine two complementary imaging modalities. This research is validated by an MCNP6 computer simulation study. The backscatter imaging system developed for this research and the experimental results of the measurements carried out using the National Physical Laboratory neutron low-scatter facility are also presented in this paper.

Neutrons can provide information related to the materials with which they interact, that is, to some extent complementary to that provided by X-rays for use in non-intrusive applications. The majority of studies to date have focused on the use of thermal neutrons with there being fewer reports concerning the use of fast neutrons despite the latter being able to penetrate deeper into substances. This record describes an investigation into the use of neutron radiation (from sources such as californium-252 and americium-beryllium) to characterize defects in the walls of oil pipelines. Both sources yield a mixed radiation field of fast neutrons and γ rays which can be collimated and directed towards the steel pipeline under study. Fast neutrons are either transmitted or reflected by the pipe wall. Since the incident radiation flux on the pipeline can be determined and the scattered neutron flux can be measured, the potential exists for an assessment of the ratio of transmitted and reflected fast neutrons to be made. If defects within the pipeline are present, the ratio may be observed to change as a result relative to that of the defect-free alternative. The scattered flux will be measured by an array of organic liquid scintillation detectors, coupled with a real-time, pulse-shape discrimination system. Both γ rays and neutrons are thus retained to provide transmission information about the pipeline sector being tested. A Monte Carlo model is used to simulate a generic pipeline section in order to understand its fast-neutron scattering response under different conditions. For instance, with the pipe filled with crude oil or partially filled with a combination of oil and gas, as well as for different types of effects to its inner structure. The results of these simulations will be presented to justify the choice of scatter arrangement, the optimum angle for the detector deployment and as evidence for the assessment of defects filled with air and for pits filled with oxide-based corrosion residues.

A variety of analytical techniques have been developed and applied to the analysis of plutonium, including α-particle spectroscopy, Hybrid K-edge densitometry, X-ray fluorescence, γ-ray spectroscopy and various embodiments of mass spectroscopy. For bulk analysis requirements of environmentally-derived substances, such as soils in the context of the assessment of plutonium in contaminated land, methods that require several stages of sample preparation are generally not suitable because such preparatory work cannot be done easily, in-situ, in the field. Further, specific localized arisings in the land, i.e., particles, might be missed by a sample-based methodology. Neutron assay is generally not sufficiently sensitive for trace plutonium assay in bulk samples because the abundance of 240Pu (favored in this regard for its spontaneous fission neutron emission) is usually too small to overcome the inherent contribution from 238U. By contrast, high-resolution, γ-ray spectroscopy offers the potential to assess relatively large sample quantities, given that a correction is made for self-shielding by the sample. Broad energy germanium (BEGe) counters offer unparalleled performance in terms of both i) energy resolution (particularly in the energy region 50 – 150 keV relevant for 241Am as a conduit for 241Pu) and ii) detection efficiency. Whilst arguably the most sensitive analytical method with isotopic sensitivity for trace plutonium is accelerator mass spectrometry, high-resolution, γ-ray spectrometry can provide an independent route for bulk assay of samples without the need for extensive sample preparation. This paper will discuss the challenges associated with these approaches drawing on data taken with soil samples from the United Kingdom.

The premise behind this research is the characterisation and integration of a unique detector system on board a submersible, remotely-operated vehicle (ROV) for the end purpose of fuel debris characterisation at Fukushima Daiichi. Currently, at Fukushima Daiichi. Whilst precise knowledge of the location of the core debris at Fukushima is not known it is commonly assumed that fuel has leaked through into the base of the pedestal below and it is suggested that it may have moved outside of the pedestal into the lower plenum. The flooding of the reactor floors immediately following the Fukushima accident adds an extra element of complexity for the detection system requiring it to be submersible and to hold any detector system in water-tight confinement. The research presented here focusses on the use of a CeBr 3 inorganic scintillator detector with a unique configuration of an in-built HV supply for ease of integration within an ROV in a submerged environment. The detector has been tested in several environments: small wave tank for source identification and a TRIGA reactor and a 60Co irradiator. It is hoped that the CeBr 3 detector will constitute one component of an on-board detector payload to determine the suitability for the localisation and identification of fuel debris inside the cores at Fukushima.

Neutron radiography is a non-destructive approach that has evolved in the last decades in contrast to the use of imaging facilities located in proximity of nuclear reactors and spallation sources. The majority of prior techniques exploit thermal neutrons by virtue of their relatively easy detectability, with there being less focus on fast-neutron imaging. In this research, we exploit fast neutrons emitted by a low-dose and mobile californium-252 source to perform fast neutron radiography. Simultaneously with neutrons, γ rays are also emitted by this particular isotope from spontaneous fission. The detector system used in our work affords the detection, real-time, of both fast neutrons and γ rays by means of pulse gradient analysis performed by a mixed field analyser. This retains event histories of both types of radiation, thus making possible to undertake, at the same time, neutron-, γ- and combined n-γ- radiography with the same field of view. The novelty of this technique is the use of a single radiation source and detection system that yields to two different radiographs of same object. In this particular record, we apply this technique to image a laptop battery, with the aim to highlight the presence of lithium ion polymer inside the battery’s cells. Our study shows how combined n-γ imaging is a valid alternative to X-rays imaging when there is the need to depict low-atomic number elements, as well as to discern elements with similar density. The system presented in this study is potentially applicable in safety & security inspection; for instance, it has the potential to be coupled with the advanced cabin baggage screening systems in airports, improving their detection performance.

This paper presents an investigation into response of instabilities of EJ-309 liquid scintillator detectors. A brief review of common instabilities associated with the photomultiplier tubes (PMTs) is presented. The energy response, energy linearity and warm-up duration of sixteen EJ-309 detectors coupled to PMTs tested is presented. A single-channel mixed-field analyser digitiser system was used for data acquisition. Furthermore, timing information of the common instability behaviours is presented alongside suggestions on how to correct for such instabilities. The results show that a single-energy energy calibration is sufficient to ensure energy linearity; the detectors must be warmed-up by ~45 minutes before stable response is achieved; the re-warm-up duration depends on the duration of the high voltage supplied to the PMT being switched off. The results indicate that the PMTs take approximately 2 hours to reach "cold" state, where a full warm-up duration must be applied. The reported instability effects will be taken into account when developing a sophisticated auto-calibration methodology for a multi-element scintillator detector system.

This paper describes the development of a submersible system based on a remote-operated vehicle coupled with radiation detectors to map the interior of the reactors at the Fukushima Daiichi nuclear power station. It has the aim oflocating fuel debris. The AVEXIS submersible vehicle used in this study has been designed as a low-cost, potentially disposable, inspection platform that is the smallest of its class and is capable of being deployed through a 150 mm diameter access pipe. To map the gamma-ray environment, a cerium bromide scintillator detector with a small form factor has been incorporated into the AVEXIS to identify radioactive isotopes via gamma-ray spectroscopy. This provides the combined system with the potential to map gamma-ray spectra and particle locations throughout submerged, contaminated facilities, such as Units 1, 2 and 3 of the Fukushima Daiichi nuclear power plant. The hypothesis of this research is to determine the sensitivity of the combined system in a submerged environment that replicates the combination of gamma radiation and water submersion but at lower dose rates.

Imaging of nuclear fuel using radiation has been carried out for decades for a variety of reasons. Two important reasons are Physical Invertory Verification (PIV) and Quality Assurance (QA). The work covered in this review focuses on the imaging of nuclear fuel using ionising radiation. The fuels investigated are both fresh and spent, composed of assorted materials, and in various physical forms. The radiations used to characterise the nuclear fuel include γ, α, β, muons, neutrons and X-rays. The research covered in this review, spans the past four decades and show how the technology has developed over that time. The advancement of computing technology has greatly helped with the progression of the images that are produced. The field began with 2D images in black and white showing the density profiles of γ rays from within an object, culminating in 2013 when a pebble bed fuel element was reproduced in 3D showing each 0.5 mm UO2 globule within it. With the ever increasing computing technology available to the industry, this can only mean an increase in the rate of development of imaging technologies like those covered in this review.

This work describes the first known the use of electrokinetic treatments and ionic salt washes to remediate concrete contaminated with 137Cs. A series of experiments were performed on concrete samples, contaminated with K+ and 137Cs, using a bespoke migration cell and an applied electric field (60 V potential gradient and current limit of 35 mA). Additionally, two samples were treated with an ionic salt wash (≤400 mol m−3 of KCl) alongside the electrokinetic treatment. The results show that the combined treatment produces removal efficiencies three times higher (>60%) than the electrokinetic treatment alone and that the decontamination efficiency appears to be proportional to the initial degree of contamination. Furthermore, the decontamination efficiencies are equivalent to previous electrokinetic studies that utilised hazardous chemical enhancement agents demonstrating the potential of the technique for use on nuclear licensed site. The results highlight the relationship between the initial contamination concentration within the concrete and achievable removal efficiency of electrokinetic treatment and other treatments. This information would be useful when selecting the most appropriate decontamination techniques for particular contamination scenarios.

The response of charge coupled devices (CCDs) to $\alpha $ radiation has been investigated to assess the potential of using these devices to perform $\alpha $-particle detection and measurement. Previous research indicates that $\alpha $ radiation leaves distinctive patterns in the images produced by CCDs when compared with other types of radiation. In this research, CCD responses to $\alpha $ and $\beta $ radiation have been compared to highlight these differences. The CCD used was a Sony ICX825AL interline transfer CCD with an $9.98 \times 6.71$ mm imaging area consisting of $1392 \times 1040$ pixels, each of size $6.45 \times 6.45 {\mu } m$, covered with a thin aluminised Mylar film to stop exposure to light. Images have been taken of exposures to a $^{\mathbf {210}}$Po source and horizontal streaks have been observed. It is thought the horizontal streaks are caused by the charge produced by the $\alpha $ particles being read out over several cycles. The average length of the streaks was found to be $14.5 \pm 2.0$ pixels. Future research will be done to further assess streaks produced by different energy $\alpha $ particles to investigate the potential of using CCDs for spectroscopy, using the MC40 cyclotron at the University of Birmingham.

This paper identifies a digitizer technology with unique features that facilitates feedback control for the realization of a software-based technique for automatically calibrating detector responses. Three such auto-calibration techniques have been developed and are described along with an explanation of the main configuration settings and potential pitfalls. Automating this process increases repeatability, simplifies user operation, enables remote and periodic system calibration where consistency across detectors’ responses are critical.

Scintillation detectors offer a single-step detection method for fast neutrons and necessitate real-time acquisition, whereas this is redundant in two-stage thermal detection systems using helium-3 and lithium-6, where the fast neutrons need to be thermalized prior to detection. The relative affordability of scintillation detectors and the associated fast digital acquisition systems have enabled entirely new measurement setups that can consist of sizeable detector arrays. These detectors in most cases rely on photomultiplier tubes, which have significant tolerances and result in variations in detector response functions. The detector tolerances and other environmental instabilities must be accounted for in measurements that depend on matched detector performance. This paper presents recent advances made to a high-speed FPGA-based digitizer. The technology described offers a complete solution for fast-neutron scintillation detectors by integrating multichannel high-speed data acquisition technology with dedicated detector high-voltage supplies. This configuration has significant advantages for large detector arrays that require uniform detector responses. We report on bespoke control software and firmware techniques that exploit real-time functionality to reduce setup and acquisition time, increase repeatability, and reduce statistical uncertainties.

A 75 MBq 252Cf neutron source stored inside a steel water tank was characterised using a compact fast-neutron and gamma-ray imaging system based on a passive slot modulation imaging technique. Radiation fields were imaged from a variety of positions with the source in the stored position (located in the center of the water tank: high shielding) and in the exposed position (located at the edge of the water tank: low shielding). It was possible to locate the 252Cf source in each image and gain additional information of the neutron and gamma-ray fields in the local environment including scatter contributions from the steel shield, floor and walls in proximity to the source. A long exposure image of the stored source, taken over thirty days, identified the location of the radiation source from the low dose ( < 1 μ Sv / h ) field penetrating through 46 cm of water in the storage tank.

This work describes the first known the use of electrokinetic treatments and ionic salt washes to remediate concrete contaminated with

In this paper the contribution of primary and secondary radiation from a water phantom to a pinhole volume, as a result of three neutron sources (Cf, AmBe and 5 MeV mono-energetic) and two gamma sources (Cs and Co), is separately estimated using the PTRAC particle tracking option available in MCNP. Also in this paper imaging of the mixed radiation field produced by a Van de Graaf accelerator (when a water phantom is present) is described. In the model, a spherical tally volume, 2 cm in diameter, was placed equidistantly from a radioactive source and 30×30×15 cm3 water phantom. Monte Carlo simulations have been carried out to investigate the level of primary and secondary radiation contributing to the pinhole volume directly from the source and from interactions in the phantom respectively. The spatial distribution of counts clearly discriminated the source and the phantom. The results have shown that the percentage of neutrons reflected from the phantom with energies above 1 MeV increases with mean energy of the source. This method has significant potential to characterise secondary radiation in proton therapy, where it would help to verify the location and the energy delivered during the treatment.

In this paper, characteristics of neutron source localisation using an EJ-426 detector with a polyethylene collimator are presented. The detector is placed inside of the polyethylene collimator, where the collimator has a constant internal diameter and external diameter gradually increases towards the back end. The combined detector-collimator unit was mounted onto an equatorial mount and, data were collected from an ultra-fast mixed-field analyser. Neutron images have been produced to characterise the area investigated in real-time. Radiation images are fundamentally related to the detector efficiency and to the collimator geometry. In order to characterise the neutron source localisation capability of the polyethylene collimator, with an EJ-426 detector, experiments have been carried out using three different neutron sources. Results indicate that the existing design of the polyethylene collimator is suitable for localisation of low energy neutron sources. The measurements were performed in the low-scatter facility of the Neutron Metrology, National Physical Laboratory, Teddington, UK.

The design, principle of operation and the results of measurements made with a four-channel organic scintillator system are described. The system comprises four detectors and a multiplexed analyzer for the real-time parallel processing of fast neutron events. The function of the real-time, digital multiple-channel pulse-shape discrimination analyzer is described together with the results of laboratory-based measurements with $^{252}{\hbox{Cf}}$ , $^{241}\hbox{Am–Li}$ and plutonium. The analyzer is based on a single-board solution with integrated high-voltage supplies and graphical user interface. It has been developed to meet the requirements of nuclear materials assay of relevance to safeguards and security. Data are presented for the real-time coincidence assay of plutonium in terms of doubles count rate versus mass. This includes an assessment of the limiting mass uncertainty for coincidence assay based on a 100 s measurement period and samples in the range 0–50 g. Measurements of count rate versus order of multiplicity for $^{252}{\hbox{Cf}}$ and $^{241}\hbox{Am–Li}$ and combinations of both are also presented.

A phantom for the evaluation of electrokinetic remediation of radioactive species from water saturated concrete is described. The phantom has been designed to be a general analogue for environments where structural concrete is saturated by radioactive aqueous solutions and where electrokinetic remediation may be deployed. It is also a specific analogue for the walls of storage ponds for legacy spend nuclear fuel pins where the pond water comprises a large volume of hazardous active waste that may penetrate the pond wall. The fabricated phantom was evaluated using a fixed electrical potential to monitor the rate of cationic transport of K+ through concrete samples of different thickness (20 and 35 mm respectively). Results of the evaluation show K+ diffusion coefficients of 5.20 × 10−13 and 7.61 × 10−13 m2 s−1 for the 20 mm and 35 mm samples, consistent with those seen in literature for the transport of caesium through concrete of similar thickness. The phantom offers a means of experimental validation of computational electrokinetic models as well as providing a basis to test the effects of electrode material on ionic transport rates, to interrogate the effects of pH on all components of the system, and as a basis for instruction, education and training in nuclear decommissioning and waste treatment.

High spin states in the isotope 194Hg were populated using the 150Nd (48Ca,4n) reaction at a beam energy of 213 MeV. A sequence of dip ole transitions has been observed above 8 MeV excitation energy. Cross-over transitions have also been identified. An interpretation connecting this sequence to a nuclear shape change is attempted. A comparison with similar structures in the neighbouring Hg isotopes is also attempted.

We report on the development of a technique for the rapid production of exfoliated graphite oxide, using an electrochemical oxidation and exfoliation method. Vitreous carbon foam (VCF) sheets were utilised as the anode in an electrolytic cell. VCF anodes were oxidized under a maximum applied electric field of 60V and a current of 2A in NaOH electrolyte solutions. The resulting solids, collected after separation of the electrolyte, were analysed using ultraviolet-visible spectroscopy, X-ray diffraction and scanning electron microscopy. The preliminary analysis shows that the solids collected were comprised of graphite oxide and partially reduced graphite oxide platelets. Quartz substrates coated in partially reduced graphite oxide solutions were thermally annealed to produce conductive layers.

The use of a thallium-doped sodium-iodide well-type scintillation detector for the assay of the low-activity radioisotope 40K, in open-source potassium chloride aqueous solutions, is described. The hazards, safety concerns and radiowaste generation associated with using open-source radioactive isotopes can present significant difficulties, the use of hot cells and escalated costs in radioanalytical laboratory research. A solution to this is the use of low-hazard alternatives that mimic the migration and dispersion characteristics of notable fission products (in this case 137Cs). The use of NaI(Tl) as a detection medium for naturally-abundant levels of 40K in a range of media is widespread, but the use of 40K as a radioactive tracer has not been reported. The use of such low-activity sources is often complicated by the ability to detect them efficiently. In this paper a scintillator detector designed to detect the naturally-abundant 40K present in potassium chloride in tracer applications is described. Examples of the use of potassium chloride as a tracer are given in the context of ion exchange and electrochemical migration studies, and comparisons in performance are drawn from literature with hyper pure germanium semiconductor detectors, which are more commonly utilised detectors in high-resolution counting applications.