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Long-lived high-level waste from commercial nuclear power reactors is a problem that concerns stakeholders and scientists working in the back end of the nuclear fuel cycle. Nuclear waste transmutation is under investigation to tackle this problem, transforming nuclides that represent a long-term source of radioactivity, radiotoxicity, and heat into short-lived or stable nuclides. However, the transmutation process will require that several long-lived isotopes be separated from the spent nuclear fuel, which raises proliferation concerns. In this paper, we perform an investigation of the attractiveness characteristics related to the material used in a lead-cooled fast reactor system concept designed to burn minor actinides before and after irradiation. The materials evaluated are separated uranium, neptunium, plutonium, americium, and curium. We also evaluated grouped product materials, neptunium + americium and neptunium + plutonium. Additionally, we present potential safeguards and physical protection implications for the proposed materials. The main conclusion of this paper is that the separated neptunium and plutonium generated by the fast reactor are materials that deserve attention mainly related to physical protection measures. [ABSTRACT FROM AUTHOR]

Uranium hexafluoride (UF6) is a significant concern for material accountancy and verification in the international safeguards community. Verification of the contents of UF6 cylinders is generally attempted with gamma spectroscopy but the current methods assume a uniform, homogeneous UF6 mass distribution within the cylinder. In this work, it was found experimentally and confirmed via modeling, that under an external heat load (the sun), the UF6 and its daughter products undergo fractionation in the cylinder. This fractionation of the UF6 and daughter products can cause an errant measurement of the enrichment of the cylinder when using the current verification methods. [ABSTRACT FROM AUTHOR]

This comprehensive review paper delves into the multifaceted aspects of nuclear safeguards, emphasizing their indispensable role in upholding global security and preventing the illicit use of nuclear materials. Offering a fresh perspective on the nuclear safeguard system, the paper conducts a thorough examination of nuclear material accounting, underscoring the pivotal role of the International Atomic Energy Agency (IAEA) in implementing and verifying these safeguards. It provides valuable insights into the safeguarding technologies, including the utilization of sophisticated Non-Destructive Assay (NDA) equipment such as neutron counting and Gamma-ray spectroscopy, which are essential for accurately characterizing nuclear materials, elucidating the crucial role of these activities in ensuring the integrity and effectiveness of nuclear safeguards. Furthermore, the study sheds light on the complexities inherent in the nuclear fuel cycle, elucidating various stages ranging from uranium mining to waste disposal. Additionally, it discusses different fuel cycle options and their implications for non-proliferation efforts. This discussion sheds light on how safeguards are integrated into each stage of the fuel cycle to prevent the proliferation of nuclear materials. The paper delves into the Additional Protocol, emphasizing its significance in enhancing the effectiveness of nuclear safeguards by providing the IAEA with expanded inspection access and information regarding nuclear activities. This comprehensive analysis not only enhances understanding of the current safeguarding landscape but also underscores the significance of ongoing research and development in this vital realm of nuclear safety and security. [ABSTRACT FROM AUTHOR]

Copyright of Journal of Isotopes is the property of Journal of Isotopes Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Gamma spectrometry analysis of uranium enrichment is an important technical tool for international nuclear safeguards verification. As a kind of special nuclear materials，235U is an important verification object of nuclear safeguards verification. In addition，the determination of uranium enrichment in samples is a key measurement for product process control，process measurement and waste characterization in uranium enrichment production facilities，uranium conversion facilities，fuel manufacturing plants and reprocessing plants，as well as a key technical means for tracking illicit trafficking in nuclear materials and responding to terrorist threats and attacks in homeland security activities. Simulated annealing algorithm is aiming to simulate the solid annealing process，and iteratively solve the solution problem by continuously reducing the annealing temperature and combining with the Metropolis criterion. The result has been proved to converge to the global optimal solution according to probability 1，which is one of the common solutions for solving optimization problems. In the article，a uranium enrichment calculation method based on simulated annealing algorithm is proposed for gamma spectrometry analysis. A corresponding mathematical model is established for the practical problem of enrichment calculation，the generation and iterative process of algorithmic solutions are improved by using the sampling theory and the forbidden searching strategy，and the iterative solution is combined with the nonlinear least-squares fitting for the enrichment of the samples，which realizes the relative efficiency of the curve fitting and the calculation of the enrichment degree within the energy range of 143 keV to 1001 keV. The performance test of the algorithm and the results of sample analysis show that the uranium enrichment calculation method based on the improved simulated annealing algorithm has the ability to escape the local optimal problem in uranium enrichment calculation，and can be applied to the enrichment calculation of different uranium samples，the relative deviation of the results of the enrichment analysis over all measured samples is within ±5%.

This comprehensive review paper delves into the multifaceted aspects of nuclear safeguards, emphasizing their indispensable role in upholding global security and preventing the illicit use of nuclear materials. Offering a fresh perspective on the nuclear safeguard system, the paper conducts a thorough examination of nuclear material accounting, underscoring the pivotal role of the International Atomic Energy Agency (IAEA) in implementing and verifying these safeguards. It provides valuable insights into the safeguarding technologies, including the utilization of sophisticated Non-Destructive Assay (NDA) equipment such as neutron counting and Gamma-ray spectroscopy, which are essential for accurately characterizing nuclear materials, elucidating the crucial role of these activities in ensuring the integrity and effectiveness of nuclear safeguards. Furthermore, the study sheds light on the complexities inherent in the nuclear fuel cycle, elucidating various stages ranging from uranium mining to waste disposal. Additionally, it discusses different fuel cycle options and their implications for non-proliferation efforts. This discussion sheds light on how safeguards are integrated into each stage of the fuel cycle to prevent the proliferation of nuclear materials. The paper delves into the Additional Protocol, emphasizing its significance in enhancing the effectiveness of nuclear safeguards by providing the IAEA with expanded inspection access and information regarding nuclear activities. This comprehensive analysis not only enhances understanding of the current safeguarding landscape but also underscores the significance of ongoing research and development in this vital realm of nuclear safety and security.

Copyright of Taiwan Law Review is the property of Angle Publishing Co., Ltd and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

According to a recently proposed nuclear safeguards technique, monitoring the power-normalized, ex-core neutron detection rate over time could be used to detect undeclared changes to the fissile composition of a reactor core. In this study, Monte Carlo simulations have been used to verify some of the underlying assumptions of this technique and the possibilities of using it to detect undeclared fuel substitutions during the first 2-year cycle of a light water small modular reactor. Depletion calculations and neutron transport simulations were used to study the changes in the power-normalized neutron leakage rate $${J_{\rm{b}}}/{P_{{\rm{core}}}}$$ J b / P core through the core barrel upon fuel substitutions and whether these changes are fully explained by changes in the core fissile composition. Several substitution scenarios have been studied, where partially depleted fuel assemblies were substituted with fresh fuel assemblies after 1 year of irradiation. The modeled substitution scenarios, which included substituting up to 4 out of 37 fuel assemblies in the core at a time, resulted in changes in $${J_{\rm{b}}}/{P_{{\rm{core}}}}$$ J b / P core of up to 3.5% depending on which fuel assemblies were substituted. The results indicate that the ex-core neutron signatures are not only sensitive to core-averaged nuclide densities, fission cross sections, and neutron flux, but also the spatial distributions of these and other parameters throughout the core. Effects such as these could mean that monitoring the core fissile composition with the proposed technique might be more complex than previously suggested. [ABSTRACT FROM AUTHOR]

AbstractUntil a long-term solution for the disposal of spent nuclear fuel (SNF) is available, interim dry casks will be increasingly used for the storage of SNF discharged from civilian nuclear power reactors. Dry casks containing commercial SNF may hold several significant quantities of plutonium, so appropriate nuclear material safeguards monitoring is needed. An external remote monitoring system (RMS) has been developed to advance dry cask safeguards monitoring beyond the current method of containment and surveillance used to maintain continuity of knowledge.In this study, neutron transport simulations of SNF assemblies in a dry cask were performed for several special nuclear material diversion scenarios. The simulations considered various loading patterns and fuel storage durations as long as 100 years. For each fuel loading pattern and storage time investigated, the simulation results were used to calculate the required measurement time to achieve a nondetection probability $\beta $β ≤ 10% for the diversion of any single fuel assembly in the cask. The calculations were performed for false alarm probabilities $\alpha $α as low as 0.0001% (or 10−6). A Monte Carlo postprocessing approach was developed to consider the impact on the required measurement time of uncertainty in the burnup of fuel assemblies.The study found that the external RMS is well suited for the surveillance of SNF in dry cask storage for nuclear safeguards or other purposes and is able to detect the diversion of a single SNF assembly even after decades of storage and with a very low false alarm probability. [ABSTRACT FROM AUTHOR]

The Republic of Korea is performing independent national inspections under the IAEA's State System of Accounting for and Control (SSAC), and developing an evaluation methodology for the material unaccounted for (MUF) to reinforce capabilities with the purpose of assessment for the accounting system of the facility handling bulk nuclear materials. In relation to this, a new approach for MUF evaluation was proposed in this study based on the guide to the expression of uncertainty in measurement (GUM). Both the conventional MUF evaluation method and the GUM method were applied to a hypothetical list of inventory items including material balance. Considering the ease of uncertainty propagation according to the GUM, it was assumed that independent uncertainty factors correspond to random factors, while correlated uncertainty factors correspond to systematic factors. The total MUF uncertainties were similar for both methods; however, it was verified that some uncertainties were affected by the measurement procedure in the GUM method. Furthermore, the GUM method was found to be more conducive to conducting a factor analysis for the MUF uncertainty. It was therefore concluded that application of the GUM approach could be beneficial in cases of national safeguard inspections where factor analysis is required for MUF assessment.

The notion of deterrence by denial has been introduced in relation to biosecurity. The goal of deterrence by denial in biosecurity extends beyond dissuading attacks. It also encompasses the protection of citizens in the event of an attack. In this vein, it may be considered as including the elements of deterrence and safeguards in terms of the protection of complying states from the hazards of violations and evasions. In presenting a historical overview of nuclear safeguards and shedding light on the limitations of the safeguard system of the International Atomic Energy Agency, this paper delineates the distinction between safeguards and verification for deterrence. Subsequently, it examines the case of the Biological Weapons Convention in which verification is considered a crucial factor for deterrence and investigates the potential for deterrence by denial to serve as an alternative to verification in effective safeguards. It argues that the adoption of deterrence by denial in biosecurity could exert synergistic effects by rectifying the shortcomings of the verification system within the Biological Weapons Convention and providing effective safeguards without the need for coercive measures.

Uranium-bearing particle analysis is an important technology in the field of nuclear safeguards. In uranium-related nuclear activities, especially in the processes of uranium enrichment, uranium-bearing aerosols are unavoidable released into the environment. There are significant differences between the uranium particles formed by uranium-bearing aerosols after drying and that from nature in terms of isotopic ratio, elemental composition, and impurity composition. Therefore, we can provide a judgment basis for the monitoring of nuclear facilities and activities by analyzing the characteristic information of micrometer or submicron sized uranium-bearing particle. After nearly 30 years of development, some relatively mature methods for the analysis of uranium-bearing particles have been developed. In this review, the process of particle analysis including sample collection, preliminary screen, particle recovery, identification and localization, and measurement is briefly introduced. The advantages and disadvantages of the commonly used technical in each process are compared. At the same time, according to the different purposes of analysis, uranium-bearing particle analysis is divided into three research directions: uranium isotope analysis, morphology and elemental composition analysis, and age-dating. Recent progress of each direction is discussed. Finally, based on the progress of the research, combined with the Development and Implementation Support Program for Nuclear Verification of the IAEA, the future research direction of uranium-bearing particle analysis technology is prospected.

Decay energy spectroscopy (DES) is an increasingly popular technique for measuring isotopic composition of actinide samples for nuclear safeguards applications. Current approaches for actinide DES utilize milligram-scale external gold absorbers (> 0.1 nJ/K) that are integrated with actinide samples through mechanical kneading and are thermally connected to microcalorimeters using indium or gold wire bonds. This leads to relatively slow sensor rise time and, consequently, limits counting speed to a few counts per second. We are developing faster metallic magnetic calorimeter-based DES by integrating actinide samples with magnetic sensor materials. This reduces signal rise time and enables high counting speed while maintaining the ability to knead the radioactive source with the absorber. We have measured signal rise time of 0.7 μs with a 1.5 mg external gold absorber using this approach. We also demonstrated online DES operation using an Ortec DSPEC 50, a commercially available data acquisition system developed for semiconductor detectors. [ABSTRACT FROM AUTHOR]

The aim of the present work was to develop a direct method for age dating (production date measurement) of uranium samples by laser ablation multi-collector inductively coupled plasma mass spectrometry (LA-MC-ICP-MS) by the measurement of the 230Th/234U ratio. The major instrumental conditions and sample characteristics affecting the accuracy and precision were investigated in this systematic study. By comparing the obtained LA-MC-ICP-MS results with those obtained after chemical separation measurements, it shows that the LA-MC-ICP-MS method is capable to produce accurate results for pure highly enriched uranium. Natural and low-enriched uranium, however, needs a higher mass resolution to remove the identified interferences, which can lead to erroneous results. [ABSTRACT FROM AUTHOR]

AbstractThe identities of unknown analytes within four eutectic LiCl-KCl melts were determined using electrochemical methods, simulating the uncertainty of electrochemically probing an electrorefiner salt bath or molten salt nuclear reactor. With a variety of electrochemical methods (e.g. cyclic voltammetry, chronopotentiometry, and square-wave voltammetry), and electroanalytical techniques (e.g. semi-differentiation), every analyte was positively identified, although one false positive occurred because of an unexpected chemical interaction. This study highlights some remaining challenges for the use of electrochemical sensors in nuclear material control and accountability in molten salts: (1) quantification of analytes without the use of calibration curves (e.g. error in property values, such as diffusion coefficient) and (2) additional and interfering electrochemical signals due to interaction and alloying of multiple species. [ABSTRACT FROM AUTHOR]

Public attention on blockchain is currently centered on the erratic fluctuation of cryptocurrency, overshadowing other potential use-cases that can have significant impact on global security, including the tracking, accounting, and securing of sensitive assets such as nuclear material and facilities. [ABSTRACT FROM AUTHOR]

This paper describes the methodology used to create a fuel data library comprising safeguards-relevant quantities that may be useful for verification of spent nuclear fuel (SNF) produced by simulating a concept Molten Salt Reactor (MSR). The Monte-Carlo particle transport code, Serpent2 and the calculation code SOURCES 4C were used to compile this fuel data library. The data library is based on the Compact Molten Salt Reactor (CMSR) concept being developed by Seaborg Technologies (based in Copenhagen, Denmark). The library includes data such as nuclide mass densities for a total of 1398 nuclides (in g/cm3), as well as total decay heat production (denoted by suffix the ‘TOT_DH’) in Watts, total gamma photon emission rates (denoted by the suffix ‘TOT_GS’) in photos per second, and the total activity (denoted by suffix ‘TOT_A’) in Becquerel. Lastly, the data also includes total neutron emission rates from 1) spontaneous fission (denoted by ‘SF’ and reported in neutrons per second per cm3), and 2) (ɑ, n) reactions (denoted by ‘AN’ and reported in neutrons per second per cm3) for the fuel salt. These quantities are reported for a range of burnup-initial enrichment-cooling time (or collectively known as, BIC) parameters. The resulting fuel data library is an extension of a previously published data library for the same reactor concept but with one significant change. The current library is based on a more realistic model of the CMSR involving movement of gaseous and volatile fission products (GFP and VFP) from the core via an Off-Gas System (OGS). The dataset is made available for public use in a compressed binary format as an HDF5 (or Hierarchical Data Format) file that can be parsed using data analysis tools such as Pandas.

Gamma-ray spectrometry uses the X-ray and characteristic gamma-rays associated with 235U, 238U and their decay daughters in uranium samples for analytical calculations to achieve rapid and timely nondestructive detection of sample enrichment. As one of the methods for determining uranium enrichment by gamma-ray spectrometry, the relative detection efficiency calibration method has been commonly used in nuclear safeguards on-site inspections to measure the enrichment of samples. Its core is the relative efficiency self-calibration using multiple characteristic gamma-rays of 235U in the low energy region of 120-186 keV and the characteristic gamma-rays of 238U decay daughter 234Pam in the high energy region of 258-1 001 keV. In this paper, the relative detection efficiency of the detector was obtained using the Monte Carlo method based on the actual characterization results of a coaxial-type high-purity germanium detector, the fitting results and errors of three different types of fitting formulas in different energy region ranges were investigated. A comprehensive analysis of the effect of the fitting formulas from the low-energy region to the medium-high energy region on the relative efficiency of the characteristic peaks shows the logarithmic transfer series function performs better results among the three fitting formulas. During the on-site inspection, the aluminum and stainless steel shieldings of the sample and the self-absorption of the uranium dioxide and uranium hexafluoride body source have a certain impact on the relative detection efficiency curve. The specific influence and changing trend of these factors on the curve were studied, and the fitted curves for thicker shields and strong self-absorption were then corrected by adding correction factors. The results show that after adding the correction factor, the fitting error of the relative detection efficiency curve for different shielding and self-absorption cases is reduced from 2%-4% to 1%-2% at the characteristic peaks used for enrichment calculation. The results of this paper have reference value for the nondestructive assay of uranium enrichment in the on-site inspection of nuclear safeguards.

The history of uranium isotopic reference materials from JRC-Geel during the last 50 years is reviewed by presenting certification methods and relevant applications. The certified isotope ratios are traceable to the SI via gravimetrical preparation, either directly through gravimetrical mixing of highly enriched materials, or indirectly using existing gravimetrically prepared reference materials for calibration of mass spectrometers used for certification measurements. Due to developments of mass spectrometers and analytical methods, certification measurements have improved regarding precision, uncertainties and accuracy. This has led to a comprehensive set of uranium isotopic reference materials available for nuclear safety, security and the scientific community. [ABSTRACT FROM AUTHOR]

Pending the availability of an operational long-term spent nuclear fuel (SNF) repository or other disposal methods, SNF will be increasingly stored in interim dry casks. Casks loaded with commercial SNF may contain several significant quantities of plutonium, so appropriate nuclear material safeguards monitoring is in order. An external remote monitoring system (RMS) developed by researchers at Texas A&M University is proposed to further the current dry cask safeguards regime, which is limited to containment and surveillance mechanisms. In this study, neutron measurements of SNF in dry cask storage were performed with the external RMS at a commercial interim spent fuel storage installation. Corresponding neutron transport simulations using MCNP were conducted with two types of detector responses (tallies) and the results were compared with measurements. The objectives of the study were to add dry cask measurement data to the literature, to assess the performance of the external RMS in full-scale dry cask measurements, and to investigate the degree to which measurements could be estimated with high-fidelity radiation transport simulations. The study demonstrated that the external RMS can acquire neutron count rate measurements with a relative error of less than 0.5% in 5 min or less through the shielding of a dry cask lid. Additionally, the developed simulation model matched trends in the measurement data to a degree that exceeds results in current literature, and normalization factors were calculated to better estimate the magnitude of neutron count rates. [ABSTRACT FROM AUTHOR]

This paper describes the creation and description of a nuclear fuel isotopics dataset for irradiated fuel salt from a Molten Salt Reactor (MSR). The dataset has been created using simulations carried out using the Monte-Carlo particle transport code, Serpent 2.1.32 (released February 24, 2021) and the calculation code SOURCES 4C (released October 09, 2002) for computing properties of irradiated molten fuel salt. The dataset comprises isotopic mass densities of 1362 isotopes (including fission products and major and minor actinides) and their corresponding contributions to decay heat, gamma activity, and spontaneous fission rates computed by Serpent 2.1.32 as well as overall neutron emission rates from spontaneous fission and (ɑ, n) reactions computed by SOURCES 4C. These quantities are computed for a model MSR core utilizing a full-core 3D model of the Seaborg Compact Molten Salt Reactor (CMSR). The dataset spans a wide range of values of burnup (BU), initial enrichment (IE) and cooling time (CT) over which the above-mentioned quantities are reported.The structure of the dataset includes isotopic mass densities (in g/cm3), followed by isotope-wise contributions to decay heat (denoted by suffix ‘_DH’ and reported in Watts), gamma photon emission rates (denoted by suffix ‘_GS’ and reported photons per second), and spontaneous fission rates (denoted by suffix ‘_SF’ and reported in fissions per second). In addition to these columns, the data also includes total neutron emission rates from 1) spontaneous fission (denoted by ‘SF’ and reported in neutrons per second per cm3), and 2) (ɑ, n) reactions (denoted by ‘AN’ and reported in neutrons per second per cm3). In total, the dataset has 310,575 rows of different combinations of fuel burnup, initial enrichment, and cooling time (BIC) values spanning the realistic possible range of these parameters. The dataset is made available for public use in a comma-separated value file that can be easily read using one of the numerous popular data analysis tools such as NumPy or Pandas.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

In a fissile material, the inherent multiplicity of neutrons born through induced fissions leads to correlations in their detection statistics. The correlations between neutrons can be used to trace back some characteristics of the fissile material. This technique, known as neutron noise analysis, has applications in nuclear safeguards or waste identification. It provides a nondestructive examination method for an unknown fissile material. This is an example of an inverse problem where the cause is inferred from observations of the consequences. However, neutron correlation measurements are often noisy because of the stochastic nature of the underlying processes. This makes the resolution of the inverse problem more complex since the measurements are strongly dependent on the material characteristics. A minor change in the material properties can lead to very different outputs. Such an inverse problem is said to be ill posed. For an ill-posed inverse problem, the inverse uncertainty quantification is crucial. Indeed, seemingly low noise in the data can lead to strong uncertainties in the estimation of the material properties. Moreover, the analytical framework commonly used to describe neutron correlations relies on strong physical assumptions, and is thus inherently biased. This paper addresses dual goals. First, surrogate models are used to improve neutron correlation predictions and quantify the errors on those predictions. Then the inverse uncertainty quantification is performed to include the impact of measurement error alongside the residual model bias. [ABSTRACT FROM AUTHOR]

Through a systematic scanning of 235U and 238U emission lines between 280 nm and 745 nm, the optimal emission line for direct gaseous uranium hexafluoride (UF6) enrichment assay using laser-induced breakdown spectroscopy (LIBS) was found. Screening for spectral features that are potentially useful for U isotopic analysis was gauged from the magnitude of the 235U–238U isotopic shift and the signal-to-background ratio of the emission line through a parameter termed ΔSBR 235U–238U. The ΔSBR spectrum shows peaks at wavelength positions where there are strong lines with significant 235U–238U shifts. The screening identified 13 spectral-window candidates, which were down selected based on their overall accuracy in predicting the 235U enrichment of three UF6 samples of natural (0.720 atom% 235U) and low-enriched (4.675 atom% and 9.157 atom% 235U) grades. The U(I) 646.498 nm emission line, with a determined 235U–238U isotopic shift of −17.7 pm, was found to be the optimal spectral window for direct UF6 enrichment assay. The root mean square error for enrichment assays on the three natural and low-enriched UF6 samples, with each sample measured in six replicates, was 0.31% in absolute 235U content. Each measurement comprised LIBS signals accumulated from 3000 laser shots. The analytical bias and precision were better than 0.5% and 0.3%, respectively, in absolute [235U/(235U + 238U)] ratios. Specific for the two low-enriched UF6 samples, the relative standard deviations from six replicated measurements were around 2%. Graphical Abstract This is a visual representation of the abstract. [ABSTRACT FROM AUTHOR]

The liquid sampling-atmospheric pressure glow discharge (LS-APGD) ionization source has proven to be an effective analysis tool for making uranium isotope ratio measurements when coupled to high-resolution mass spectrometers, such as the Orbitrap. While previous studies have shown the capabilities of the LS-APGD for isotope ratio determination, a systematic evaluation of the measurement uncertainty of the technique has not been conducted. To this end, the International Standards Organizations (ISO) guidelines to the expression of uncertainty in measurement (GUM analysis) have been applied to generate an uncertainty budget. Presented here, a preliminary assessment derived from the GUM analysis was performed. The uncertainty in the instrument blank determination has been identified as a primary factor contributing to measurement uncertainty for the LS-APGD-Orbitrap method. These findings for the specific test case of uranium isotopic analysis will be invaluable in applications across the breadth of isotope ratio mass spectrometry performed on this unique instrumental platform. [ABSTRACT FROM AUTHOR]

In the framework of the European Commission Support Programme to the International Atomic Energy Agency (EC SP task A1753) 20 samples were obtained from the Dolní Rožínka (DIAMO, Czech Republic) uranium milling facility. The sampling procedure followed stepwise the uranium production and purification from the U ore to uranium ore concentrate (yellow cake) end-product. Elemental concentrations, rare-earth elemental pattern, anion concentrations, morphology and isotope abundance ratios of S, Sr, Pb and U were measured at each sampling stage. The purpose of the measurements was to investigate the applicability of various material characteristics for authentication, propagation and variation of these parameters, and to identify the relevant signatures for nuclear forensics and safeguards during the uranium production. [ABSTRACT FROM AUTHOR]

The International Atomic Energy Agency uses the Thermal Ionization Mass Spectrometry Total Evaporation method routinely to analyze the n(U-235)/n(U-238) atom amount ratios in inspection and quality control samples. This paper uses a top-down statistical model applied to quality control measurements of certified reference materials (CRM) to study the capabilities of the technique. The technique is shown to produce measurements with a relative standard deviation not more than 0.03% when applied to a wide range of certified reference materials. [ABSTRACT FROM AUTHOR]

Copyright of Yıldırım Beyazıt Law Review (YBLR) is the property of Ankara Yildirim Beyazit University Law School and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

1 1 Notice: This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (https://www.energy.gov/doe-public-access-plan). Safeguards verification measurements of nuclear material content in fresh fuel salt for liquid-fueled molten salt reactors (MSRs) are likely to be required as part of nuclear material accountancy for International Atomic Energy Agency safeguards. This paper presents a comprehensive review and evaluation of 18 potential candidate techniques to quantify total uranium and 235 U for input accountancy measurements for liquid-fueled MSRs. As part of an overall screening and down-selection effort to identify the most promising techniques for further development for an MSR feed monitoring system, this paper defines eight figures of merit (FOMs): reasonably achievable measurement uncertainty, measurement time required, capital cost, burden upon the facility operator, maintenance intensity, technological maturity, human capital requirements for operation, and whether the technique introduces a path for potential material removal. Each candidate technique is then evaluated across these FOMs to identify the techniques with the highest potential for future development for fresh fuel accountancy measurements in MSRs. Our findings indicate that no single technique or combination thereof currently has the requisite technological maturity for immediate implementation in nuclear material accountancy at a liquid-fueled MSR facility. While several promising techniques are identified, there is a critical lack of experimental data for most systems in the context of molten salt applications. • We propose criteria to screen feed accountancy techniques in molten salt reactors. • We evaluated 18 candidate techniques to identify the most promising for development. • Presently, no technique is sufficiently mature for MSR feed accountancy measurements. • Critical gaps in measurement data of actinide-bearing molten salts must be addressed. [ABSTRACT FROM AUTHOR]

We introduce an active learning framework for content-based image retrieval for video surveillance that can be trained ad-hoc for a single camera in a matter of minutes. This technique allows searching for both, known and unknown objects, given a region of interest. The process does not require prior labelled data and treats image retrieval as a binary classification task, in which frames can be similar or different from a query image. The technique is compatible with any pre-trained deep feature extractor. In addition, we propose a novel label propagation algorithm that benefits from (1) visual similarity of image pairs and (2) the semantic representation of the feature vectors from a pre-trained deep feature extractor. This approach allows to reduce the amount of labels needed, while avoiding the propagation of errors. Our experiments with three use-cases from a nuclear facility show the validity of the proposed method, which achieves high precision and recall while requiring minimal amounts of labelled data. [Display omitted] • Present a relevant video frame retrieval pipeline that can be trained interactively. • Propose an active learning algorithm that leverages visual and semantic similarity. • Visual similarity makes label propagation less prone to introducing errors. [ABSTRACT FROM AUTHOR]

Copyright of Mural Internacional is the property of Editora da Universidade do Estado do Rio de Janeiro (EdUERJ) and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

With the nuclear efforts of North Korea in the news, along with the ups and downs of the Iran agreement, there have been more and more articles in the popular press about nuclear inspections, safeguards, and verification of nuclear weapons agreements. But what are the nuts and bolts behind how one goes about ensuring that a nuclear deal is being adhered to? How do inspectors know if a nuclear program is peaceful or not -- especially without unfettered access to every single last part of all the components of the program? Here, the author describes how the International Atomic Energy Agency pulls together the various pieces of the puzzle to achieve a comprehensive picture of a state's nuclear program and detect any anomalies that might hint of ambitions towards nuclear weaponry. [ABSTRACT FROM AUTHOR]

Dry casks will be a prevalent spent nuclear fuel (SNF) storage option until solutions for long-term storage or disposal are deployed. A dry cask storing 32 pressurized water reactor fuel assemblies will likely contain about 20 significant quantities of plutonium, so these structures require effective safeguards monitoring. An external remote monitoring system (RMS) is proposed to advance the current dry cask safeguards regime which relies on containment and surveillance. The objectives of this study were to assess the performance of the external RMS as a detection system and to develop a simulation approach for estimating measurements. Small-scale experiments of generic neutron source diversions mimicking SNF diversion from a dry cask were conducted and the nondetection probability was calculated for a variety of measurement times. MCNP simulations were carried out to assess the degree to which the measurement results could be predicted. A previous simulation methodology was advanced to consider uncertainty in the activity of sources being measured. The study concluded that the external RMS performs well as a neutron detection system and that MCNP simulation is a viable tool both for predicting measurements made with the external RMS and for calculating nondetection probabilities of hypothetical, generic diversion scenarios. [ABSTRACT FROM AUTHOR]

• A partial defect detector design proposed for spent fuel inspection before transfer from storage pool to difficult-to-access storage facility. • Performance indicator of a detector applied to the spent fuel inspection was quantitatively analyzed from a regulatory standpoint. • Detection capability of new design was assessed with the randomly distributed fuel defect scenarios. The current nuclear safeguards approach to spent nuclear fuel inspection at nuclear power stations is based on item counting and limited partial defect analysis. With the expected surge in spent fuel storage, limited spent fuel storage pool capacity, and the increasing need for transferring fuel to long-term storage facilities, there is a growing demand for more efficient and cost-effective nuclear safeguards approaches for nuclear materials management in civilian nuclear power facilities. This study proposes a scintillator-based indirect gamma detector for spent fuel inventory screening inspection, specifically designed for use in interim storage pools prior to fuel transfer to difficult-to-access storage. This paper presents the design of the proposed detector, its application for spent fuel screening inventory inspection, and analysis using MCNP for partial defect detection. Results of analysis indicated that verifying a ∼ 13.6 % level of randomly distributed fuel defect for the Westinghouse 17x17 fuel assembly is possible using this approach. The performance evaluation also indicates that inspection of spent fuel assemblies of various vendor types against 1 SQ diversion may be possible. [ABSTRACT FROM AUTHOR]

As the IAEA has applied integrated safeguards and a state level approach to member states, the importance of national inspection has increased. However, the requirements for national inspection for some member states are different from the IAEA safeguards. In particular, the national inspection for the ROK requires on-site U concentration analysis due to a domestic notification. This research proposes an on-site U concentration analysis (OUCA) method for UO2 pellets using gamma spectroscopy to satisfy the domestic notification requirement. The OUCA method calculates the U concentration of UO2 pellets using the measured net X-ray counts and declared 235U enrichment. This research demonstrates the feasibility of the OUCA method using both MCNP simulation and experiment. It simulated and measured the net X-ray counts of different UO2 pellets with different U concentrations and 235U enrichments. The simulated and measured net X-ray counts were fitted to polynomials as a function of U concentration and 235U enrichment. The goodness-of-fit results of both simulation and experiment demonstrated the feasibility of the OUCA method.

• Addressing the inverse depletion issue is essential for analyzing and comprehending the utilization and history of spent nuclear fuel. • Once-through ANN based approach is proposed to eliminate the need for the inversion step, directly connecting fuel inventory to origins. • The proposed approach has a significantly lower computational cost and higher accuracy than the benchmark approach. This work addresses the challenge of identifying the origins and burnup history of used nuclear fuel, which is crucial for safeguarding and non-proliferation analysis. It presents a comparative evaluation of two distinct computational approaches to bundles at the same in-core position: the first entails building a forward Artificial Neural Networks (ANN) surrogate model for fuel inventory prediction followed by solving the inverse depletion problem using an optimization solver. The second approach is to build an inverse ANN model that relates the used nuclear fuel inventory to its initial and burnup conditions (a once-through approach), eliminating the need for the inverse optimization solver step which results in better accuracy, less computational costs, and renders the application of such approach practical. KENO VI depletion model is utilized to deplete a VVER assembly model generating the necessary datasets for the ANN inverse model training and validation. The inverse model is then employed to estimate the initial composition and the burnup for several test cases. The results are then compared to the known initial and burnup conditions as well as to the estimations made by benchmark results from an existing ANN forward surrogate model combined with the Particle Swarm Optimization (PSO) solver. Comparative analyses across eight scenarios consistently reveal that the once-through ANN inverse model has superior accuracy and consistency in terms of estimating the actual burnup and initial fuel enrichment with a relative root mean squared error (RMSE) of 1.20 × 10 - 2 compared to the benchmark approach (ANN-PSO) which yielded a relative RMSE of 2.18 × 10 - 2 . Moreover, the once-through inverse approach has a significantly lower computational cost as it requires 1 inverse model run compared to the 9,450,000 model runs required by the benchmark approach (ANN-PSO). Overall, the results demonstrate that the proposed once-through approach simplifies and enhances the efficiency of solving the inverse fuel depletion problem while outperforming the benchmark approach in terms of predictive accuracy computational cost. [ABSTRACT FROM AUTHOR]

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