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International audience; Nuclear metrology is an essential aspect to consider for further improvements of the initial characterization of sites under decommissioning. The H2020 Euratom project INSIDER in June 2017 aims at improving the management of contaminated materials arising from decommissioning and dismantling (D&D) operations by proposing an integrated methodology for radiological characterization. This methodology is based on advanced statistical processing and modelling, coupled with adapted or innovative measurement methods. A metrological approach supports the qualification of this integrated methodology with a concrete application to real projects representative of the use cases identified in the project. Assessment of the outcomes will be used for providing recommendations and guidance resulting in pre-standardization texts.

[EN] With the aim of demonstrating the concrete advantages that novel technologies such as Virtual (VR) can provide to the nuclear industry, the authors of this paper have been working on the development of a VR based simulator of a gamma dose rate detector for training purposes, to be applied in the field of nuclear security and safety. Historically in nuclear science, simulating gamma dose rate transport has had a series of requirements, most importantly the accuracy of the computation. When embedding this dose rate computation in the environment of a VR based application, a second and opposing key requirement appears: real time performance. Meeting this requirement is only possible if a fast method to compute gamma radiation is used. In order to achieve this target the authors have been working in ways of improving the efficiency of the Point-Kernel method by reducing its computational effort. This paper presents the latest step in this pursuit of efficiency; a novel method based on a non-regular kernel approach, combined with a KD-tree based volume division method. Devised to reduce as much as possible the number of points that represent the volume of the source while aiming at retaining sufficient dose computation accuracy. (C) 2016 Elsevier Ltd. All rights reserved., This project is fully funded by the Institute of Transuranium Elements of the European Commission's Joint Research Centre, Ispra site, Italy.

Nuclear safeguards applications require accurate fission models that exhibit prompt neutron anisotropy. In the laboratory reference frame, an anisotropic neutron angular distribution is observed because prompt fission neutrons carry momentum from fully accelerated fission fragments. A liquid organic scintillation detector array was used with pulse shape discrimination techniques to produce neutron-neutron cross-correlation time distributions and angular distributions from spontaneous fission in a 252 Cf, a 0.84 g 240 Pu eff metal, and a 1.63 g 240 Pu eff metal sample. The effect of cross-talk, estimated with MCNPX-PoliMi simulations, is removed from neutron-neutron coincidences as a function of the angle between detector pairs. Fewer coincidences were observed at detector angles near 90°, relative to higher and lower detector angles. As light output threshold increases, the observed anisotropy increases due to spectral effects arising from fission fragment momentum transfer to emitted neutrons. Stronger anisotropy was observed in Cf-252 spontaneous fission prompt neutrons than in Pu-240 neutrons.

Safeguarding nuclear material is an important and challenging task for the international community. One particular safeguards technique commonly used for uranium assay is active neutron correlation counting. This technique involves irradiating unused uranium with (α, n) neutrons from an Am–Li source and recording the resultant neutron pulse signal which includes induced fission neutrons. Although this non-destructive technique is widely employed in safeguards applications, the neutron energy spectra from an Am–Li sources is not well known. Several measurements over the past few decades have been made to characterize this spectrum; however, little work has been done comparing the measured and theoretical spectra of various Am–Li sources to each other. This paper examines fourteen different Am–Li spectra, focusing on how these spectra affect simulated neutron multiplicity results using the code Monte Carlo N-Particle eXtended (MCNPX). Two measurement and simulation campaigns were completed using Active Well Coincidence Counter (AWCC) detectors and uranium standards of varying enrichment. The results of this work indicate that for standard AWCC measurements, the fourteen Am–Li spectra produce similar doubles and triples count rates. The singles count rates varied by as much as 20% between the different spectra, although they are usually not used in quantitative analysis, being dominated by scattering which is highly dependent on item placement.

A dual-particle imaging (DPI) system, capable of simultaneously imaging fast neutrons and gamma rays, has been operated in the presence of mixed-oxide (MOX) fuel to assess the system׳s angular resolution and material-characterization capabilities. The detection principle is based on the scattering physics of neutrons (elastic scattering) and gamma rays (Compton scattering) in organic and inorganic scintillators. The detection system is designed as a combination of a two-plane Compton camera and a neutron-scatter camera. The front plane consists of EJ-309 liquid scintillators and the back plane consists of interleaved EJ-309 and NaI(Tl) scintillators. MCNPX-PoliMi was used to optimize the geometry of the system and the resulting prototype was built and tested using a Cf-252 source as an SNM surrogate. A software package was developed to acquire and process data in real time. The software was used for a measurement campaign to assess the angular resolution of the imaging system with MOX samples. Measurements of two MOX canisters of similar isotopics and intensity were performed for 6 different canister separations (from 5° to 30°, corresponding to distances of 21 cm and 131 cm, respectively). The measurements yielded a minimum separation of 20° at 2.5 m (86-cm separation) required to see 2 separate hot spots. Additionally, the results displayed good agreement with MCNPX-PoliMi simulations. These results indicate an angular resolution between 15° and 20°, given the 5° step. Coupled with its large field of view, and its capability to differentiate between spontaneous fission and (α,n) sources, the DPI system shows its potential for nuclear-nonproliferation applications.

New organic–plastic scintillation compositions have demonstrated pulse-shape discrimination (PSD) of neutrons and gamma rays. We present cross-correlation measurements of 252 Cf and mixed uranium– plutonium oxide (MOX) with the EJ-299-33 plastic scintillator. For comparison, equivalent measurements were performed with an EJ-309 liquid scintillator. Offline, digital PSD was applied to each detector. These measurements show that EJ-299-33 sacrifices a factor of 5 in neutron–neutron efficiency relative to EJ-309, but could still utilize the difference in neutron–neutron efficiency and neutron singleto-double ratio to distinguish 252 Cf from MOX.

Neutron measurements can be used to distinguish plutonium in metal or oxide form, a capability that is of great interest in nuclear nonproliferation, treaty verification, and other applications. This paper describes measurements performed on well-characterized samples of plutonium oxide and plutonium metal using the pulse-shape-discrimination-capable plastic scintillator EJ-299-33. Results are compared to those obtained with a same-sized detector cell using the liquid scintillator EJ-309. The same optimized, digital pulse shape discrimination technique is applied to both detectors and the neutron pulse height distributions are compared. Results show that the EJ-299-33 plastics can be successfully used for plutonium measurements, where the gamma ray to neutron detection ratio is much higher than for typical radioactive sources. Results also show that EJ-299-33 detectors can be used to characterize plutonium samples, specifically to discriminate between plutonium metal and oxide.

Measurements were performed at the Joint Research Centre in Ispra, Italy to field test a fast-neutron multiplicity counter developed at the University of Michigan. The measurements allowed the assessment of the system׳s photon discrimination abilities, efficiency when measuring neutron multiplicity, ability to characterize 240Pueff mass, and performance relative to a currently deployed neutron coincidence counter. This work is motivated by the need to replace and improve upon 3He neutron detection systems for nuclear safeguards applications.

As new reprocessing techniques and fuel forms are developed, the ability of inspection agencies and facility operators to measure powders containing several actinides becomes increasingly necessary. Neutrons emitted from induced and spontaneous fission of different nuclides are very similar, making it difficult to measure these powders with nondestructive assay techniques. To measure the powders, a neutron multiplicity technique based on first-principle methods was developed to exploit isotope-specific nuclear properties, such as energy-dependent fission cross sections and neutron-induced fission multiplicity. This technique was tested through measurements using an epithermal neutron multiplicity counter with two different interrogation (α,n) sources and varying plutonium materials. To complement these measurements, extensive Monte Carlo N-Particle eXtended (MCNPX) simulations were performed for each measured sample, as well as samples that were not available to measure. The primary application of...

Uranium is present in most nuclear fuel cycle facilities ranging from uranium mines, enrichment plants, fuel fabrication facilities, nuclear reactors, and reprocessing plants. The isotopic, chemical, and geometric composition of uranium can vary significantly between these facilities, depending on the application and type of facility. Examples of this variation are: enrichments varying from depleted (~0.2 wt% 235U) to high enriched (>20 wt% 235U); compositions consisting of U3O8, UO2, UF6, metallic, and ceramic forms; geometries ranging from plates, cans, and rods; and masses which can range from a 500 kg fuel assembly down to a few grams fuel pellet. Since 235U is a fissile material, it is routinely safeguarded in these facilities. Current techniques for quantifying the 235U mass in a sample include neutron coincidence counting. One of the main disadvantages of this technique is that it requires a known standard of representative geometry and composition for calibration, which opens up a pathway for potential erroneous declarations by the State and reduces the effectiveness of safeguards. In order to address this weakness, the authors have developed a neutron coincidence counting technique which uses the first principle point-model developed by Boehnel instead of the “known standard” method. This technique was primarily tested through simulations of 1000 g U3O8 samples using the Monte Carlo N-Particle eXtended (MCNPX) code. The results of these simulations showed good agreement between the simulated and exact 235U sample masses.

A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutrons to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials.

Methods for the determination of the subcritical neutron multiplication of nuclear materials are of interest in the field of nuclear nonproliferation and safeguards. A series of measurements were performed at the Joint Research Center facility in Ispra, Italy to investigate the possibility of using a time-correlated pulse-height (TCPH) analysis to estimate the sub-critical multiplication of nuclear material. The objective of the measurements was to evaluate the effectiveness of this technique, and to benchmark the simulation capabilities of MCNPX-PoliMi/MPPost. In this campaign, two low-multiplication samples were measured: a 1-kg mixed oxide (MOX) powder sample and several low-mass plutonium–gallium (PuGa) disks. The measured results demonstrated that the sensitivity of the TCPH technique could not clearly distinguish samples with very-low levels of multiplication. However, the simulated TCPH distributions agree well with the measured data, within 12% for all cases, validating the simulation capabilities of MCNPX-PoliMi/MPPost. To investigate the potential of the TCPH method for identifying high-multiplication samples, the validated MCNPX-PoliMi/MPPost codes were used to simulate sources of higher multiplications. Lastly, a characterization metric, the cumulative region integral (CRI), was introduced to estimate the level of multiplication in a source. However, this response was shown to be insensitive over the range of multiplications of interest.

This paper proposes a simplified model built with Matlab to compute an approximation of the neutron and photon flux in a BWR spent fuel assembly in a pond. Starting from a reduced number of Monte Carlo simulations, a set of transmission probabilities has been calculated. These values allow the estimation of the particle flux inside a fuel assembly placed in a generic storage rack. This calculation is possible provided the transmission probability has been calculated for the desired value of burnup and cooling time of the assemblies in the rack. The validity of the model has been tested first with a uniform pattern of the storage rack and finally with a generic configuration. Reference values from the equivalent Monte Carlo simulations show very good agreement and the relative error between the two measures is always within few percents. The main advantage of this simplified model is the great reduction of computational time required to perform the estimation and the fact that the basic idea behind the model can be applied to any fuel geometry and type of storage pond.

Radiation Portal Monitors (RPMs) are used to detect illicit trafficking of nuclear or other radioactive material concealed in vehicles, cargo containers or people at strategic check points, such as borders, seaports and airports. Most of them include neutron detectors for the interception of potential plutonium smuggling. The most common technology used for neutron detection in RPMs is based on 3 He proportional counters. The recent severe shortage of this rare and expensive gas has created a problem of capacity for manufacturers to provide enough detectors to satisfy the market demand. In this paper we analyse the design of typical commercial RPMs and try to optimise the detector parameters in order either to maximise the efficiency using the same amount of 3 He or minimise the amount of gas needed to reach the same detection performance: by reducing the volume or gas pressure in an optimised design.

This paper describes the use of the Monte Carlo code MCNPX-PoliMi for nuclear-nonproliferation applications, with particular emphasis on the simulation of spontaneous and neutron-induced nuclear fission. New models for the outgoing neutrons and gamma rays emitted in spontaneous and induced fission are described. For spontaneous fission, the models include prompt neutron energy distributions that depend on the number of neutrons emitted in the individual fission events. For neutron-induced fission, due to lack of data, the prompt neutron energy distributions are independent of the number of neutrons emitted in the individual fission events. Gamma rays are sampled independently of the neutrons. Code validation is performed on well-characterized mixed-oxide fuel and plutonium-oxide samples.

Detection of illicit trafficking of nuclear material relies on the detection of the radiation emitted. In the case of plutonium, one of the characteristic signatures derives from neutron emission. For this reason, neutron detectors cover an important role in detection systems. Most current neutron detection systems used for nuclear security are based on the 3He technology. Unfortunately, in the last few years the market of 3He has encountered huge problems in matching the supply and the demand. The need has grown significantly due to the increasing demand of instrumentation for security. This has caused an exponential increase of the price from one side and on the other side a serious strategic problem of resources. In order to guarantee the availability of detection systems for nuclear security, it is necessary to develop alternative detection systems based on technologies different from 3He. Many research projects have been devoted for the development of novel neutron detectors both by research organisations and by industries. Scientists from the PERLA laboratory of the Joint Research Centre (JRC) in Ispra, Italy, and their collaborators have tested several of these novel concepts in the last couple of years. This paper describes the detector systems tested at JRC and preliminary results on detectors that can be considered as promising alternatives to 3He.

This paper describes the calibration, characterisation and Monte Carlo modelling of a new IAEA Uranium Neutron Collar (UNCL) for LWR fuel, which can be operated in both passive and active modes. It can employ either 35 3 He tubes (in active configuration) or 44 tubes at 10 atm pressure (in its passive configuration) and thus can be operated in fast mode (with Cd liner) as its efficiency is higher than that of the standard UNCL. Furthermore, it has an adjustable internal cavity which allows the measurement of varying sizes of fuel assemblies such as WWER, PWR and BWR. It is intended to be used with Cd liners in active mode (with an AmLi interrogation source in place) by the inspectorate for the determination of the 235 U content in fresh fuel assemblies, especially in cases where high concentrations of burnable poisons cause problems with accurate assays. A campaign of measurements has been carried out at the JRC Performance Laboratories (PERLA) in Ispra (Italy) using various radionuclide neutron sources ( 252 Cf, 241 AmLi and PuGa) and our BWR and PWR reference assemblies, in order to calibrate and characterise the counter as well as assess its performance and determine its optimum operational parameters. Furthermore, the fast-UNCL has been extensively modelled at JRC using the Monte Carlo code, MCNP-PTA, which simulates both the neutron transport and the coincidence electronics. The model has been validated using our measurements which agreed well with calculations. The WWER1000 fuel assembly for which there are no representative reference materials for an adequate calibration of the counter, has also been modelled and the response of the counter to this fuel assembly has been simulated. Subsequently numerical calibrations curves have been obtained for the above fuel assemblies in various modes (fast and thermal). The sensitivity of the counter to fuel rods substitution as well as other important aspects and the parameters of the fast-UNCL performance have been investigated and assessed.

When applying numerical simulation to physical problems it is important to be able to clearly understand the quality that can be reached and the uncertainty associated with the modelling. In this paper we will deal with the Monte Carlo simulation of neutron coincidence and multiplicity counters. These instruments are intensively used in nuclear safeguards for the measurement and verification of the mass of fissile materials in nuclear installations. Their numerical simulation, either for detector design or for numerical calibration, is now used more and more frequently as support to inspections. Several benchmark exercises have been organised in order to assess the capabilities of Monte Carlo simulation to correctly predict realistic practical cases in neutron coincidence and multiplicity counting. Nevertheless most of them allowed reaching conclusions only on the overall uncertainties associated to the modelling through the comparison between calculated and measured integral data. In this paper we aim to analyse more deeply the sources of these uncertainties, separating the different possible causes and analysing parametrically the individual factors. We will include influencing factors deriving from the physical model of the detector, of the acquisition electronics, of the sample and from the nuclear data. The final total uncertainties are consistent with the results of the integral benchmarks.

International safeguards inspectorates (e.g., International Atomic Energy Agency {IAEA}, or Euratom) rely heavily on neutron assay techniques, and in particular, on coincidence counters for the verification of declared nuclear materials under safeguards and for monitoring purposes. While 3He was readily available, the reliability, safety, ease of use, gamma-ray insensitivity, and high intrinsic thermal neutron detection efficiency of 3He-based detectors obviated the need for alternative detector technologies. However, the recent decline of the 3He gas supply has triggered international efforts to develop and field neutron detectors that make use of alternative materials. In response to this global effort, the U.S. Department of Energy’s (DOE) National Nuclear Security Administration (NNSA) and Euratom launched a joint effort aimed at bringing together international experts, technology users and developers in the field of nuclear safeguards to discuss and evaluate the proposed 3He alternative materials and technologies. The effort involved a series of two workshops focused on detailed overviews and viability assessments of various 3He alternative technologies for use in nuclear safeguards applications. The key objective was to provide a platform for collaborative discussions and technical presentations organized in a compact, workshop-like format to stimulate interactions among the participants. The meetings culminated in a benchmark exercisemore » providing a unique opportunity for the first inter-comparison of several available alternative technologies. This report provides an overview of the alternative technology efforts presented during the two workshops along with a summary of the benchmarking activities and results. The workshop recommendations and key consensus observations are discussed in the report, and used to outline a proposed path forward and future needs foreseeable in the area of 3He-alternative technologies.« less

Fight against illicit trafficking of nuclear material relies on the possibility to detect nuclear material concealed in vehicles, people or cargo containers. This is done by equipping and training law enforcement and security staff in border stations or other points of access to strategic places and critical infrastructures with radiation detection equipment. The design, development, testing and evaluation of these instruments ideally require the use of real nuclear material to assess, verify and certify their detection performance. Availability of special nuclear material may be an issue, especially for industry, since only few specialized laboratories are licensed for such material. This paper tries to analyse and describe the possibility to use suitable surrogates that may replace the use of real nuclear material in testing the detection capabilities of instruments used in nuclear security.

The breeding potential in the irradiation channels of research reactors is of safeguards concern, because of lacking continuous supervision on the type of experiments in all the irradiation channels. Moreover, the irradiation time can be optimized in order to breed high quality weapon grade plutonium. With regard to the safeguards measures currently adopted, IAEA concentrates its efforts on those reactors whose thermal power is greater than 25 MWth, because it was calculated that a 25 MWth LEU-fuelled reactor produces not more than one Significant Quantity of Pu (8 kg)/year in its spent fuel and a HEU-fuelled reactor of this power would require an annual reload of not more than one Significant Quantity of U235 (25 kg). In order to investigate whether it would be possible to determine an analogous power level threshold to estimate the clandestine plutonium production capability of different research reactors, the Monte Carlo method was used to determine the neutron flux in the irradiation channels and to calculate the plutonium breeding potential for three different reactor types: (1) a Triga Mark II with 250 kWth, representative for a small size research reactor; (2) a Material Test Reactor (MTR) with 5 MWth, representative for a medium size research reactor; (3) a High Flux Reactor (HFR) with 45 MWth, representative for a large size research reactor. It was observed that the most important factors for plutonium breeding are the neutron flux (to which reaction rates are proportional) and the available space to place irradiation samples. The breeding capability scales fairly well with the reactor power level and from about 10 MWth onwards the proliferation concern raises with increasing power level and available sample space.

This paper describes the principle of load cell monitoring in Gas Centrifuge Enrichment Plants (GCEPs) and how this technique can be implemented in order to improve nuclear safeguards in these kinds of installations. We present a few different possibilities of exploitation of the data that can be obtained by acquiring in continuous mode the weights of the cylinders in the feed and withdrawal (F/W) stations and how to derive conclusions concerning the operation of the plant (conformity or not with respect to “normal” operation). Furthermore the different diversion scenarios for GCEP are discussed and we show how mass monitoring techniques can contribute to detect each of them. Finally we resume the advantages and limitations of the technique and outline some important practical issues concerning the implementation of process monitoring in GCEP, in particular for what concerns the inspectorate/operator relationships. Load cell monitoring could play an important role in the safeguard implementation of future GCEPs.

The knowledge of a neutron energy spectrum is extremely valuable in nuclear non-proliferation and safeguards applications. Unfortunately, conventional gas and fission-type detectors are not adapted to determine incident neutron energy. As a result, a matrix inversion technique was implemented to generate desirable and consistent neutron energy spectra from an EJ-309 liquid scintillator. To overcome the ill-posed unfolded solutions, the method of sequential least squares is reviewed briefly, and the conditions under which it is used are stated. From the analysis, unfolded energy spectra solutions of five distinct PuO2 fuel samples and a 252 Cf source are presented and compared to the simulated unfolded and reference energy distributions.

The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is in this case taken similar to the source of the Power Plant Conceptual Study – Water Cooled Lithium Lead (PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanket the first wall of the ITER design is attached to a molten salt zone, in which ThF 4 and UF 4 solute salts are transported by a LiF–BeF 2 solvent salt. For this blanket design, the fissile material is assessed in quantity and quality for both the Th-U and the U-Pu fuel cycle. The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket are simulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated with the burn-up code ORIGEN-S. For the Th-U cycle the bred material output remains below 10 g/h with a 232 U impurity level of 30 ppm, while for the U-Pu cycle supergrade material is produced at a rate up to 100 g/h.

Neutron coincidence counting is the reference NDA technique used in nuclear safeguards to measure the mass of nuclear material in samples. For high-enriched uranium (HEU) samples active neutron interrogation is generally performed and the most common device used by nuclear inspectors is the Active Well Coincidence Counter (AWCC). Within her master thesis at the Polytechnic of Milan, the first author performed an intensive study on the characteristics and performances of the AWCC in order to assess the 235 U mass in HEU oxide samples at the PERLA laboratory of JRC. The work has been summarised in this paper that starts with the optimisation of the use of AWCC for nuclear safeguards, describing the calibration procedure, reporting results of a series of verification measurements, summarising the performances that can be obtained with this instruments during inspections at fuel production plants and concluding with the discussion of uncertainties related to these measurements.

We present new results on neutron and gamma-ray pulse-height distributions (PHDs) measured with liquid scintillators from five plutonium-oxide samples of varying mass and burnup and a 252 Cf isotopic source. We show that the analysis of the pulse-height distributions can be used to easily distinguish the fissile material (plutonium oxide) from the 252 Cf source. Moreover, the slope of the measured pulse-height distributions can be analyzed to distinguish the burnup of the samples, independent of their masses. Finally, the measurement scenarios are modeled using the MCNP-PoliMi code and good agreement is observed between the measured and the simulated pulse-height distributions.

Neutron coincidence counting is a widely adopted nondestructive assay (NDA) technique used in nuclear safeguards to measure the mass of nuclear material in samples. Nowadays, most neutron-counting systems are based on the original-shift-register technology, like the (ordinary or multiplicity) Shift-Register Analyser. The analogue signal from the He-3 tubes is processed by an amplifier/single channel analyser (SCA) producing a train of TTL pulses that are fed into an electronic unit that performs the time- correlation analysis. Following the suggestion of the main inspection authorities (IAEA, Euratom and the French Ministry of Industry), several research laboratories have started to study and develop prototypes of neutron-counting systems with PC-based processing. Collaboration in this field among JRC, IRSN and LANL has been established within the framework of the ESARDA-NDA working group. Joint testing campaigns have been performed in the JRC PERLA laboratory, using different equipment provided by the three partners. One area of development is the use of high-speed PCs and pulse acquisition electronics that provide a time stamp (LIST-Mode Acquisition) for every digital pulse. The time stamp data can be processed directly during acquisition or saved on a hard disk. The latter method has the advantage that measurement data can be analysed with different values for parameters like predelay and gate width, without repeating the acquisition. Other useful diagnostic information, such as die-away time and dead time, can also be extracted from this stored data. A second area is the development of “virtual instruments.” These devices, in which the pulse-processing system can be embedded in the neutron counter itself and sends counting data to a PC, can give increased data-acquisition speeds. Either or both of these developments could give rise to the next generation of instrumentation for improved practical neutron-correlation measurements. The paper will describe the rationale for changing to the new technology, give an overview of the hardware and software tools available today and a feedback of the experience gained in the first tests. Associated with the experimental tests, the ESARDA-NDA working group is also performing an intercomparison benchmark exercise on the analysis software for pulse processing.

For the first time, measurements of the time-dependent cross-correlation distributions of plutonium oxide have been made separately for neutrons and gamma rays. Six EJ-309 liquid scintillation detectors with a digital, offline pulse shape discrimination and pulse timing method were used to measure five different samples of varying mass and burnup. The number of (neutron, neutron) correlations were selectively analyzed versus plutonium mass and a clear, increasing trend was observed. Additionally, the measurement scenarios were modeled using the MCNP-PoliMi code and good agreement was observed between the measured and simulated cross-correlation functions.

A paper describing the conceptual idea of using passive neutron assay for the verification of large size uranium samples in fuel fabrication plants was first presented at the 2001 ESARDA conference. The advantages of this technique, as a replacement of active interrogation using the PHOto-Neutron Interrogation Device (PHONID) device, were evident provided that a suitable detector with higher efficiency than those commercially available would be realised. The previous paper also included a feasibility study based on the experimental data. To implement this technique, a high-efficiency passive counter (HEPC) has been designed by the JRC, Ispra. JRC has also built a first smaller-scale prototype. This paper will describe the tests made in the PERLA laboratory and report the performance of the prototype. In parallel, the design of the large HEPC has been finalised for Euratom safeguards. Two units for the fuel fabrication plants in Dessel (B) and Juzbado (E) have been produced by a commercial manufacturer under JRC specifications. The two detectors have been installed in the two sites in summer 2004 after an extensive test campaign in PERLA. Since then they are in use and some feedback on the experience gained is reported at the end of this paper.

MCNP-PTA is a new Monte Carlo code for the simulation of neutron counters for nuclear safeguards applications developed at the Joint Research Centre (JRC) in Ispra (Italy). After some preliminary considerations outlining the general aspects involved in the computational modelling of neutron counters, this paper describes the specific details and approximations which make up the basis of the model implemented in the code. One of the major improvements allowed by the use of Monte Carlo simulation is a considerable reduction in both the experimental work and in the reference materials required for the calibration of the instruments. This new approach to the calibration of counters using Monte Carlo simulation techniques is also discussed.

In this paper we discuss the long-term perspective of the application of numerical modelling to neutron (coincidence or multiplicity) counting. The advantage of this technique for nuclear safeguards is discussed with particular emphasis to the numerical calibration of detectors. As a particular case of numerical modelling, real-time simulation would allow overcoming some problems related to pre-calibration of neutron counters, especially in all those cases when the sample characteristics deviate from the calibration conditions. Nevertheless it requires a significant increase in computing capabilities. The requirements for real-time simulations deriving from inspection needs are discussed and compared to current (and possible future) computing capabilities.

This paper describes a methodology for a complete and correct assessment of the errors coming from the uncertainty of each individual component on the final result. A general methodology accounting for all the main sources of error (both type-A and type-B) will be outlined. In order to better illustrate the method, a practical example applying it to the uncertainty estimation for a special case of multiplicity counter, the SNMC developed at JRC, will be given.

The EURopean Illicit TRAfficking Countermeasures Kit (EURITRACK) inspection system uses 14 MeV neutrons produced by the D(T,n α) reaction to detect explosives in cargo containers. Fast-neutron-induced reactions inside the container produce gamma rays, which are detected in coincidence with the associated alpha particle. The definition of the neutron path and the time-of-flight measurement allow positioning the source of the gamma ray inside the container, while the chemical composition of the target material is correlated with the energy spectrum of the coincident gamma rays. However, in case of dense cargo, neutron moderation and photon attenuation inside the container make difficult the reconstruction of the material composition from the measured gamma-ray energy spectrum. An analytical method has been developed and validated against experimental data, which allows obtaining the chemical carbon-to-oxygen and carbon-to-nitrogen ratios of the inspected items from the gamma-ray energy spectra. The principle of the method is presented along with validation tests.

The EURopean Illicit TRAfficing Countermeasures Kit (EURITRACK) project is part of the 6th European Union Framework Program. It aims at developing a Tagged Neutron Inspection System (TNIS) to detect illicit materials, such as explosives and narcotics, in cargo containers. Fast neutron induced reactions produce specific gamma-rays used to determine the chemical composition of the inspected material. The associated particle technique is employed to precisely locate the interaction points of the neutrons. A new deuterium– tritium neutron generator has been developed, including a pixelized alpha particle detector. The TNIS also comprises high-efficiency fast neutron and gamma-ray detectors, a dedicated front-end electronics and an integrated software to entirely drive the system and automatically process the data. Most components have been integrated during last months at Institute Ruder Boskovic, in Zagreb, Croatia. An overview of the TNIS and of its preliminary performances is presented.

In this paper, we perform a sensitivity study on the influence of various physical and operational parameters on the results of neutron multiplicity counting. The purpose is to have a better understanding of the importance of each component and its contribution to the measurement uncertainty. Then we will be able to determine the optimal conditions for the operational parameters and for detector design and as well to point out weaknesses in the knowledge of critical fundamental nuclear data.

JRC has developed the Scrap Neutron Multiplicity Counter (SNMC): an advanced neutron multiplicity counter for the verification of inhomogeneous Pu samples, such as scrap material in MOX fuel fabrication plants. The innovative features of this counter with respect to existing ones rely on two aspects: (i) an optimised design based on Monte Carlo calculations in order to select the most appropriate materials, geometry and detector disposition for maximum efficiency and (ii) novel electronics based on digital signal processing (DSP) reducing the system dead time. The paper recalls the design process, the electronics, the construction and assembly of the counter. Then the results of the first experimental tests will be reported. We will show the characterization of the main physical parameters of the counter, the calibration and the verification of a wide variety of plutonium bearing samples available in the PERLA laboratory at JRC Ispra. This will include pure homogeneous samples (Pu dioxide powders, metal Pu, MOX powders and pellets) and some tests on heterogeneous samples representative of scrap material.

This paper aims to provide an answer to the question: is it possible to detect the diversion of a fuel assembly from a Castor container by means of passive neutron measurements? Monte Carlo simulations performed on suitably validated models suggest a correct way to tackle the problem with success perspectives. Under the assumption of ideal boundary conditions, that is neglecting the effects related to the presence of other canisters and of the environment surrounding the cask, we have demonstrated that it is possible to establish a sort of “fingerprint” of a sealed Castor cask. This fingerprint could be used for future verification of the integrity of the cask in order to re-establish the continuity of knowledge in case of failure of the C/S measures. In particular we have shown that it is possible to detect the removal of a PWR bundle from a Castor V21/A cask by means of passive neutron measurements, only if the removed bundle was originally located in peripheral positions of the fuel basket grid.

Conference of 4th International Conference on Advancements in Nuclear Instrumentation Measurement Methods and their Applications, ANIMMA 2015 ; Conference Code:121554; International audience; The Scintilla FP7 project is ended in December 2014, the fruitful results of 3 years development and tests will be presented. SCINTILLA offers the capacity to finding a reliable alternative to Helium-3 based detection systems since the gas which is predominantly used in nuclear safeguards and security applications has now become very expensive, rare and nearly unavailable. SCINTILLA benchmarks results are based on international standards. Radiation Portal tests were carried out at the Joint Research Centre (JRC) in Ispra (Italy). The scintilla project addresses few main issues. The first is to develop neutron detectors for Radiation Portal Monitor (RPM) and the second is the need of new wearable integrated solutions for Spectrometric Personal Radiation Monitor (SPRM). The partners which provide technical systems of the scintilla project are INFN-ANSALDO, CEA, SYMETRICA and SAPHYMO. For RPM, the objective is to find reliable alternatives to Helium-3 historical neutron detector and provide technical solutions which cope with tests for reliable mobile and cost effective. For Spectrometric Personal Radiation Monitor (SPRM), SCINTILLA is innovating in technology areas that offer complementary capabilities for detecting and identifying gamma Two CZT (Cadmium Zinc Telluride) addressing contexts of used by first responder technologies, one is a wearable detector and the second is a gamma camera complemented by advanced image processing technologies.

This paper starts with an historical overview of the different outreach initiatives and correspondent funding schemes supported by the European Commission in the field of nuclear security. Then it analyses the results of past collaboration under the TACIS and IfS cooperative projects, presents the status of the follow-up programme being implemented and gives an overview of the JRC expertise which can be utilized in a new nuclear safeguards and security programme, as well as under the Centre of Excellence initiative.

This paper describes the characterisation of a set of 3 He proportional counters differing in gas composition and pressure. Each step of the signal generation in the detector is analysed. Optimal working conditions are defined in order to choose the high voltage to be applied. Dead time with short shaping time is measured in such optimal conditions. In addition, the overall neutron efficiency and the gamma sensitivity are reported. Such characterisation is required for a correct choice of the associated counting electronics and the applications of the counters. A complete pulse processing modular system, consisting of a pulse height analysis system and a pulse correlation analysis system equipped with a time of arrival recorder, is used for the characterisation.

The improvement in detection limits for low and high activity samples measured with the Compton-suppression γ-spectrometer installed at the Institute for Transuranium Elements (ITU) for environmental monitoring of radioactivity, as well as nuclear safeguards, is discussed. The advantage of using two parallel acquisition lines for simultaneous measurement with and without Compton-event suppression is outlined with respect to cascade and non-cascade γ-emitters. The background reduction by Compton-event suppression made it possible to detect small peaks, which otherwise would not have been found in a conventional spectrum. In Compton-event suppression mode, the detection limit for 137Cs was improved by a factor of about 3, for 241Am we found a factor of 1.2 both in high and low active samples. The measurements of environmental reference samples showed good agreement with certified values in both acquisition modes. The application of this instrument for the determination of fission products in smear samples is described. In particular, for nuclides like 54Mn, 125Sb and 144Ce, an improvement in the detection limits by a factor of 1.6–2.4 was obtained.

The dual-particle imaging system being developed at the University of Michigan was used at the Joint Research Centre in Ispra, Italy for measurements on samples of special nuclear material. A 1,150-g mixed-oxide (MOX) fuel sample was measured with various shielding configurations to determine how the presence of lead and/or polyethylene shielding degrades the system's ability to localize a source by simultaneous neutron and photon imaging. Three two-hour measurements were taken with the source shielded by an: a) 8-mm lead sheath and 5.1-cm lead bricks, b) 8-mm lead sheath and 6.5-cm polyethylene bricks, and c) 8-mm lead sheath, 5.1-cm lead bricks, and 6.5-cm polyethylene bricks. The 8-mm lead sheath was used in all cases to reduce the measured photon count rate. A “bare” measurement was also made by using only the 8-mm lead sheath, but the unexpected presence of additional sources has rendered the measurement unsuitable for comparison. The resulting images show that the dual-particle imaging system is able to accurately localize the MOX canister in the presence of intervening material.

A dual-particle imager with sensitivity to both neutrons and photons has been developed for the detection of special nuclear material. This system has the capability to provide spectral information on detected sources that is useful for safeguard applications. In nuclear facilities with high photon background rates, neutron detection can aid in identification and verification capabilities. The dual-particle imager creates a reconstructed neutron spectrum by correlating neutron counts in pairs of liquid scintillators. The reconstructed spectrum, a combination of energy deposited and time-of-flight, contains resolution effects that blur the true, emitted source-spectrum. To reduce the impact of resolution effects, the reconstructed spectrum can be unfolded to create an estimated source spectrum. A statistical technique, maximum-likelihood expectation-maximization, has been shown to produce an unfolded result for Cf-252 and Am-Be. This technique requires a system response-matrix that contains reconstructed spectra for single neutron energies. Since measuring mono-energetic neutrons is difficult, the system-response-matrix is simulated. Monte Carlo modeling with MCNPX-PoliMi has been show to produce an accurate system response. This work expands on previous neutron unfolding results by showing results for mixed-oxide fuel and plutonium metal measured with the dual-particle imager. These results also demonstrate the ability of the dual-particle imager to discrimination between spontaneous fission neutrons and (α,n) reaction neutrons.

A plutonium metal sample was measured by a fast-neutron multiplicity counter for characterization of spontaneous fission neutron anisotropy and for verification of MCNPX-PoliMi calculations. Accurate neutron angular distribution models are important to properly simulating fast neutron coincidence measurements for nuclear nonproliferation and safeguards applications. A majority of prompt neutrons are emitted from fully accelerated fission fragments; those neutrons carry momentum from the fission fragments, and thus an anisotropic neutron angular distribution is observed in the laboratory reference frame. The fast-neutron multiplicity counter was used with pulse shape discrimination techniques to produce neutron-neutron cross-correlation time distributions from spontaneous fission in a lead-shielded 0.84 g 240Pueff metal sample. Due to neutron anisotropy, the number of observed neutron cross-correlations varied as a function of angle between a detector pair and fission source. Fewer neutron correlations were observed at detector angles near 90 degrees, relative to higher and lower detector angles. Both the neutron coincident detections as a function of time difference and detector pair angle are compared with MCNPX-PoliMi calculations and show good agreement.

The IABAT project - Impact of Accelerator Based Technologies on Nuclear Fission Safety - started in 1996 in the frame of 4(th) Framework Programme of the European Union, R&D specific programme Nucl ...

A neutron coincidence counter intended for the verification of cans of PuO2 powder has been designed and optimised using the Monte Carlo radiation transport simulation code–pulse train analyser (MCNP–PTA) which simulates both the neutron transport and the coincidence electronics setup. Severe operational and physical constraints had to be taken into account in designing and optimising the counter as well as in adopting the most adequate strategy for its calibration, validation and application. r 2007 Elsevier B.V. All rights reserved.

Conference of 2013 3rd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and Their Applications, ANIMMA 2013 ; Conference Code:102802; International audience; For Homeland Security, enhanced detection and identification of radioactive sources and nuclear material has become of increasing importance. The scintilla project aims at minimizing the risk of radioactive sources dissemination especially with masked and shielded material. SCINTILLA offers the capacity to finding a reliable alternative to Helium-3 based detection systems since the gas which is predominantly used in nuclear safeguards and security applications has now become very expensive, rare and nearly unavailable. SCINTILLA benchmarks will be based on international standards. Radiation Portal testing being carried out at the Joint Research Centre (JRC) in Ispra (Italy).