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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 the Monte-Carlo particle transport code, Serpent 2.1.32 (released February 24, 2021) and the calculation code SOURCES 4C (released October 09, 2002). 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 photos 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.

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., Order of authors in the list of papers of Vaibhav Mishra's thesis: V Mishra, E Branger, S Grape, Zs Elter, S Mirmiran

When analyzing and assessing properties of spent nuclear fuel (SNF) such as radionuclide inventories, the power history of the fuel during its time spent inside the reactor core plays an important role. This information can be very useful in the field of nuclear safeguards wherein a safeguards inspector can use it to verify the fuel properties such as burnup, initial enrichment and cooling time (or collectively termed as the “BIC” set of variables). However, such information may often be unavailable to the safeguards inspector or the level of detail in the available information may be lacking. Therefore, when analyzing SNF for various purposes (such as for safety, safeguards and back-end purposes), the power history of the fuel is most often disregarded altogether and the inspectors only look at the fuel BIC. If the power history-level information is considered, it is not uncommon to make simplifying assumptions about how the fuel is burned in the reactor. In this work, we perform an exploratory analysis of fuel cycle data from two PWR units of the Ringhals nuclear power plant in Sweden. The said analysis describes the variation in the number of cycles, cycle lengths, downtimes et cetera in order to develop a simplified yet representative model of irradiation that may be used to construct synthetic data libraries. Furthermore, we look into impact of changes in the power history on the nuclide inventories of key gamma emitters and isotopes responsible for decay heat in the SNF of the fuel in three different irradiation scenarios. Our results show that in most cases with fuels that are considerably long-cooled, it is acceptable and even preferred to use a simplified power history over an idealized or a representative irradiation model obtained from exploratory analysis of the fuel cycle data. However, for short-cooled fuels, using a simplified or even an idealized history is less preferable over modeling the detailed power history of the fuel due to the presence of sh

A majority of the nuclear reactors in the world use low-enriched uranium as fuel, and will produce plutonium during operation. This will happen when neutrons undergo capture in U238 instead of causing fission of U235, which is a likely reaction as low-enriched uranium is composed of >95% U238. While the plutonium can be used as fuel in the reactor, it is also a material highly desired by states producing nuclear weapons. Not all reactors produce plutonium of the same grade, which significantly impacts its usability in a nuclear weapon. For this reason, certain reactor technologies have been favored for military plutonium production. Heavy-water moderated reactors is one such family, that has been used in current or defunct nuclear weapons programmes by states such as India, Pakistan, Sweden, Switzerland and the United States. A number of factors impact the rate and grade of plutonium production in a reactor. These include(but are not limited to) fuel design specifications (pellet radius, fuel density), operational temperature of the fuel as well as the coolant and moderator, and numerous other operational parameters such as specific power, cycle lengths and downtimes et cetera. The present study will look into investigating the relative sensitivity of plutonium production rate and grade towards these design and operational parameters. Based on the results from this evaluation, it is expected that we can better understand which parameters impact plutonium production quantity and quality the most. This will also help us understand the role and impact of uncertainties in these parameters and connect them to the plutonium content in the spent fuel produced by these reactors

Several historic and current nuclear weapons programs are based on heavy-water moderated reactors for plutonium production. While details about the military reactors is scarce, much more information exist on civilian heavy-water moderated reactors. The CANDU reactor is the most developed one, and there is much openly available design and operation data that can be used to simulate it. The CANDU design uses heavy water for moderation and cooling together with low-enriched uranium fuel which is frequently refueled. This reactor type is not optimized for military plutonium production, but is on the contrary extensively used in civil energy production. It has never been unambiguously proven that a CANDU reactor has been used for producing plutonium for nuclear weapons. In the case of both India and Pakistan, another type of reactor is believed to have been used to manufacture plutonium. This type is based on the NRX (National Research Experimental) reactor at Chalk River Laboratories in Canada, a predecessor of the CANDU reactor. The NRX was moderated by heavy water but cooled by light water. It used natural metal uranium as fuel and was for a time the most powerful research reactor in the world. NRX was in operation between 1947 to 1993, and suffered a partial meltdown accident in 1952, which led to a substantial development in reactor safety. In this poster we are comparing the CANDU and NRX designs and discuss how the fuel handling and operation differ. The results of our simulations allow comparison of parameters such as plutonium production, the plutonium vector, throughput, and can be used for further work to assess the production at military reactors of an unknown design. The objective of this work is to gain a better estimate of the plutonium production at the reactors at Khushab in Pakistan.

The Differential Die-away Self Interrogation (DDSI) instrument was researched for many years under the Next Generation Safeguards Initiative Spent Fuel effort. Later a prototype instrument was manufactured and used to make non-destructive measurements of spent nuclear fuel in the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Sweden in 2018. Results of DDSI research, based on either simulations or measurement time, have indicated that the instrument could successfully be used to draw safeguards-relevant conclusions about spent nuclear fuel.In this work we investigate how well the modelled response of the DDSI instrument, based on Serpent and MCNP simulations, corresponds to measured data of 17x17 pressurised reactor fuel. We also studied repeatability, i.e. to what extent repeated measurements on the same fuel assembly gave consistent results. We also investigated the dependence of tau on the selected time window. The results show that tau values determined from measurement data are consistently higher than tau values determined from simulations, and that the magnitude of tau is dependent on the choice of time window. We also note that tau is relatively insensitive to positioning in the DDSI instrument.

The “Swedish line” was an ambitious program in the 1950s-1960s aiming to make Sweden self-sufficient with respect to nuclear technology. The plan was to combine civil power production with military plutonium production, should Sweden decide to develop nuclear weapons. The program included, among other things, domestic uranium mining, domestic uranium fuel production, operation of heavy-water reactors and a plutonium laboratory. The Ågesta nuclear power plant just outside Stockholm was part of the program. The reactor, also known as R3/Adam, was in operation in 1964-1974. The reactor was an underground, heavy-water cooled and heavy-water moderated pressurized reactor, providing district heat and a modest amount of electricity to the near-by suburb Farsta. After being closed down, the heavy water was sold to Canada while the fuel and some other equipment were removed. Large parts of the facility were preserved for several decades, but is now undergoing decommissioning, a process which is planned to be finalised in 2025. Within this project we are investigating how to locate and reconstruct historic information on the operation of the Ågesta reactor. Of particular interest is the nuclear fuel and its irradiation history in the reactor. There are many reasons for this: i) information and knowledge management about the operation of a nuclear facility under the “Swedish line”, ii) for assessments regarding to the produced plutonium qualities and quantities and iii) for nuclear safeguards verification of Ågesta fuel before encapsulation and final storage. For operating reactors, information about the fuel and its irradiation in the reactor is typically kept with the operator and follows the fuel as it undergoes transport and spent fuel management. In this case, the information is not straightforward to access, as the information has been distributed among multiple actors, and because custody of various parts of the information has changed over time. In this poster we

The Digital Cherenkov Viewing Device (DCVD) is one instrument available to authority inspectors for verifying spent fuel assemblies in wet storage. The measurements result in images of the Cherenkov light emissions from the fuel assembly under study. This work presents research on applying image analysis and statistical methods to improve data quality and to extract more information from the measurements, extending the use of these methods beyond what is currently implemented in the DCVD software. The goal of this project is to apply template matching and statistical analysis to the images. However, before such techniques can be applied, effort is needed to ensure that the measurements are directly comparable. Two main issues are investigated here, the first being the positioning of the Region-Of-Interest. By developing an automated Region-Of-Interest placer, a consistent and reproducible Region-Of-Interest placement can be achieved. The second is automatic identification of fuel type, to support a later comparison with a template. We demonstrate that a method based on Principal Component Analysis can be used to determine the fuel type. Finally, we present the first results regarding template matching, comparing a measured image to a template, aiming to identify regions in the image where the two differ. Such differences could be due to a partial defect located in that region, but also due to other reasons such as debris covering the fuel top. Automatically identification of such regions can in the future be used to focus inspector attention to features requiring expert judgement, supporting efficient use of the measurement data and inspector effort. The first results demonstrate the feasibility of the method, but also that more work is required before the method is robust.

The Digital Cherenkov Viewing Device (DCVD) is one of the instruments used by safeguards inspectors to verify spent nuclear fuel in wet storage. The DCVD can be used for partial defect verification, where the inspectors verify that 50% or more of an assembly has not been diverted. The methodology is based on comparing the measured Cherenkov light intensity with a predicted intensity, calculated with operator information. Recently, IAEA inspectors have encountered fuel assemblies for which systematic deviations between predictions and measurements could be observed, indicating that the prediction model did not take into account all sources of Cherenkov light production. One contribution to the Cherenkov light intensity that is frequently omitted is the contribution from beta decays, where energetic electrons exit the fuel material and enter the water with sufficient energy to directly produce Cherenkov light. The objective with this work was hence to study beta contributions and evaluate whether that could be the cause of discrepancy between predictions and experimental data. By simulating the beta contribution for fuel assemblies where the discrepancy was experimentally observed, it was determined that beta decays were the cause. The fuel assemblies had fuel rods with relatively small radii, thin cladding, a short cooling time and an irradiation history that resulted in a relatively large beta contribution for assemblies that had a comparatively low burnup. Therefore, the beta contribution was significant, and caused 10-40% of the total Cherenkov light intensity. By including the beta contributions in the predictions, the RMSE of the deviation between prediction and measurement could be reduced from 20.7% to 11.6% for the available measurement data. The results highlight that the beta contribution can be significant and should be taken into account for accurate predictions.

The Alva Myrdal Centre for nuclear disarmament (AMC) was established in 2021. AMC consists of six working Groups, and one of them - Working Group 4 - is called Technical nuclear non-proliferation and safeguards. This is the State-of-the-Art Report of that working group. The objective with the report is to provide an overview of the technical fields relevant to the working group and to highlight where research and activities within the working group may contribute to global nuclear disarmament. The report gives a brief explanation of actors in the field, introduces nuclear materials and assay techniques, and then continues to elaborate on challenges and needs associated with nuclear measurements and assessments in the fields of non-proliferation, nuclear safeguards and nuclear disarmament. A section is also devoted to the management of nuclear weapons materials after disarmament. Lastly, the report contains a section on interdisciplinary research and development in nuclear disarmament, and information about technical education and training in the non-proliferation and disarmament field., Alva Myrdal Centre for nuclear disarmament

The majority of the energy in a nuclear explosion is released in the immediate blast and the initial radiation accounts. The remaining fraction is released through radioactive decay of the explosion's fission products and neutron activation products over a longer time span. This allows for the detection of a nuclear explosion by detecting the presence of residual decay. Radionuclide monitoring stations for detection of radioactive emissions to the atmosphere is thereby an important tool in the verification of compliance with nuclear disarmament treaties. In particular, the globally spanning radionuclide station network of the International Monitoring System (IMS) has been implemented for verification of the Comprehensive Nuclear-Test-Ban Treaty. High Purity Germanium (HPGe) detectors are workhorses in radionuclide monitoring. The detection of characteristic gamma rays can be used to disclose the presence of signature nuclides produced innuclear weapon tests. A particular development that has potential to improve the sensitivity of radionuclide monitoring is the coincidence technique where decaying nuclides that emit several coincident gamma rays can be detected at much smaller activity concentrations than with conventional gamma spectroscopy. In this project, dedicated gamma-gamma coincidence detectors are being developed, utilizing electronically segmented HPGe detectors. These detectors are expected to be highly sensitive to low-activity samples of nuclides that present coincident emissions of gamma rays. In this paper we present the concept, define performance parameters, and explore the performance of such detectors to a subset of radionuclides of particular CTBT relevance. In addition, we discuss the path forward in developing a next generation gamma-gamma coincidence spectroscopy system of segmented HPGe.

Under the AMC, a range of activities covering education, research and outreach are foreseen. One of them concerns education and the build-up of competence related to disarmament, and for that reason collaborative efforts have been ongoing during 2021 and 2022 to develop a PhD-level course in verification of nuclear test explosions, and to offer it during September-October 2022. The course has developed by Uppsala University and the Swedish Defence Reserach Agency (FOI) and corresponds to 7.5 credits. It is a cross-disciplinary course that spans over several disciplines. It introduces the participants to treaties and verification regimes governing nuclear weapons and it explains identification, calculation and analysis of signatures from nuclear weapon explosions. Furthermore, effort has been made to let the participants actively work with data collection, aggregation, analysis and with the interpretation and evaluation of data. The course includes also both a laboratory exercise on detection of radionuclides, and a project work in which the participants analyze a test explosion scenario and summarize their findings and conclusions in a manner very similar to how this is done in reality. This poster will describe the details of the course and its content. Since the course is planned to be offered just before this conference, we also hope to provide some information on its execution, as well as feedback from the participants.

As a result of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT), the International Atomic Energy Agency conducts nuclear safeguards verification to ensure accurate accounting for and control of nuclear materials in states. Experts have highlighted a loophole in the NPT, whereby safeguarded nuclear material could be removed and used for military applications such as naval propulsion purposes. The loophole is exemplified by the announcement by the AUKUS security pact that the United Kingdom and United States will help Australia acquire nuclear-powered submarines. Naval reactors have so-far only been operated by nuclear-weapon states and non-signatories to the NPT. This situation puts various nuclear non-proliferation issues in the spotlight. The main concern is that Australia, a non-nuclear weapon state, may set a precedent to remove nuclear material from its civil nuclear fuel cycle; an act that could inspire non-nuclear weapon states to use naval propulsion programs as a cover to develop nuclear weapons. As a first step in addressing these concerns, we have been studying naval reactor cores and their properties and that work will be presented here. Openly available information has been used to create a model of a naval reactor core and its fuel. The model has been implemented in the Monte Carlo code Serpent2, in which also the operation of the reactor has been modelled. The objective is to study the evolution of the fuel material composition over time and to make an assessment of how useful the material is for nuclear weapon production.

The interest of the scientific community in alternate nuclear fuel cycles such as that of the Molten Salt Reactors (MSRs) has grown and waned in the past. In the recent decades however, there has been a resurgence of interest in them owing to the ever-rising demand for sustainable energy and growing concerns of climate change. However, these reactors are yet to enter the stage of commercial operation. The MSR family form a broad spectrum of reactor concepts with unique fuel cycles, designs, online reprocessing options et cetera and some of these have progressed from the phase of conceptualization to deployment as experimental reactors. Since MSRs differ significantly from the traditional Light Water Reactors (LWRs), both conceptually and from an operational point of view they come with new and unique challenges. On the issue of safeguards, the main challenges arise from the lack of experience in the industry on handling bulk fuel material such as highly radioactive and corrosive molten fuel salts containing nuclear material in place of traditional fuel items like spent nuclear fuel (SNF) assemblies. There is also a marked lack of measurement techniques and instruments that could be brought into use for enforcing safeguards on molten salt systems. There are novel ideas that have been proposed to increase the proliferation resistance of the fuel cycle in MSRs and deter the misuse of fuel material.Among the key MSR designs that have been developed at the Oak Ridge National Laboratory (ORNL) in the past was the Molten Salt Demonstration Reactor (MSDR). The MSDR concept was a 750 MW successor to the much smaller Molten Salt Reactor Experiment (MSRE) that was once operated at ORNL. While the MSDR was never built, its design, operation, safety, security and safeguards-related issues were studied extensively at several national labs in the United States. In the current study, we aim to model the irradiation of fuel salt used in the MSDR. The reactor specifications are used

The Digital Cherenkov Viewing Device (DCVD) is one of the instruments available to IAEA inspectors to verify spent nuclear fuel in wet storage. The DCVD can be used for partial defect verification, verifying that 50% or more of a fuel assembly has not been diverted. The partial defect verification relies on a comparison between measured and predicted intensities, based on operator fuel declarations. Recently, IAEA inspectors have encountered spent fuels with short cooling times where there were systematic differences between predictions and measurements. Through the Swedish support program, this deviation was investigated, by studying various modelling assumptions that could cause the discrepancy. The predominant cause of the discrepancy was beta-decay electrons, passing through the fuel cladding and entering the water with sufficient energy to directly produce Cherenkov light. Analysis of measurement data for a set of fuels where the discrepancy was found to be pronounced revealed that for modern fuel designs with thin claddings the beta contribution is enhanced, and for short-cooled fuels additional beta-decaying isotopes are abundant and must be considered. Furthermore, the data showed that for nuclear fuels that had not reached the discharge burnup, the fuel irradiation history may cause a relative enhancement of the abundance of beta-decaying isotopes relative to other isotopes causing Cherenkov light. Other studied modelling assumptions, such as void, burnable absorbers and using binned gamma spectra, showed that they only introduced a modest bias, and proper default values and data handling can mitigate it. A method to predict the direct beta contribution to the Cherenkov light intensity was developed, which can ensure that the observed biases will be eliminated from future verification campaigns. It is advised that this enhanced prediction method be included in the DCVD software, and made available to inspectors.

Plutonium and uranium are well known weapons-usable nuclear materials which are currently placed under international safeguards. Examples of nuclear-safeguards methods are surveillance of nuclear facilities, inspections of spent nuclear fuel and process monitoring at fuel reprocessing facilities. Once material comes under safeguards, records of the amount of material are kept throughout its life cycle, and material accountancy verification is regularly performed. Deviations from the expected material balances could indicate diversion of nuclear material. It has been known for many years that there exist materials which are currently not under full international safeguards, but which could at least theoretically be used to manufacture a nuclear explosive device. One material that has attracted particular attention in this context is neptunium, which can be found in spent nuclear fuel. Although it is unclear if neptunium has ever been used in a nuclear explosive device, measures for preventing un-declared separation of neptunium have been implemented since 20 years by the International Atomic Energy Agency. However, it has never been placed under safeguards to the same extent as uranium and plutonium. In part, this decision was motivated by the relatively limited amount of neptunium available 20 years ago. Many things have happened in the nuclear industry since then, and there might be a need for re-visiting the issue. In this paper, the relationship between neptunium and international safeguards will be discussed in the context of recent trends in the nuclear energy sector and efforts to close the nuclear fuel cycle.

The MYRRHA facility, which is under development at SCK CEN in Mol, Belgium, will be a multi-purpose research facility centred around an accelerator-driven reactor cooled by lead-bismuth eutectic. The MYRRHA core will contain MOX fuel with an initial plutonium content of 30%, and the lead-bismuth coolant will result in a fast neutron spectrum in the reactor. Due to reasons such as these, the radionuclide composition of spent fuel from MYRRHA can be considerably different from that of spent fuel from typical light water reactors. These differences will affect the choice of non-destructive verification techniques to safeguard the spent fuel. Further practical considerations, for example considering spent-fuel storage in lead-bismuth eutectic, will affect the choice of measurement technique. In a project led by Uppsala University, first studies of the requirements on gamma-ray and neutron measurements of spent fuel from MYRRHA have been performed using depletion calculations and Monte Carlo simulations. These studies have included i) an investigation into the radionuclide composition and gamma-ray and neutron emission in spent fuel from MYRRHA, ii) an analysis of the possibilities to discriminate between spent MYRRHA fuel and spent light-water reactor fuels using gamma-ray and neutron signatures and machine learning, iii) a feasibility study of using gamma-ray spectroscopic measurement for assaying a MYRRHA fuel assembly stored in lead-bismuth eutectic, and iv) a feasibility study of using neutron measurements under similar conditions. In the paper, new results connecting these studies will be summarised to highlight some important consequences for future safeguards verification at fast reactor systems.

Pakistan performed at least two nuclear weapons tests in 1998 as a direct response to the Indian nuclear tests earlier the same year. With this act, Pakistan became the seventh country to successfully complete a nuclear weapons programme. The Pakistani nuclear weapons arsenal consists of both uranium and plutonium weapons and the country has an extensive nuclear industry with all facilities necessary for enrichment of uranium, production of plutonium and reprocessing of spent reactor fuel. Pakistan acquired a Canadian civil heavy water nuclear reactor in 1971; KANUPP-1. In 1976 however, the cooperation with Canada ended as Canada stopped supplying fuel for the reactor. At this point, Pakistan had acquired know-how and experience to manufacture its own fuel and also started building an independent nuclear industry with several unsafeguarded reactors at the Khushab site. With French assistance, a reprocessing plant was constructed and consequently, Pakistan is today in possession of all components necessary for developing and employing both uranium and plutonium nuclear devices. In this presentation, we will explore technical challenges associated with bringing a country such as Pakistan under the existing or proposed treaty verification following treaties such as the NPT, TPNW and FMCT. Using a simulation framework and estimates based on known physical quantities and derived abilities, we will discuss what conclusions can be drawn with regards to uranium and plutonium stockpiles.

In 2020, the Swedish government announced the intent to start up a national competence centre on nuclear disarmament in Sweden. The goal was to highlight the importance of nuclear disarmament issues, and to promote research, teaching and policy support on topics relevant to nuclear disarmament. During the spring semester 2021, the Alva Myrdal Centre (AMC) on nuclear disarmament was established at Uppsala University. The AMC combines competences from different disciplines such as peace and conflict research, applied nuclear physics, and international law, and organises the work into six different working groups. One of the working groups is focusing on technical aspects, while the remaining five working groups are focusing on policy aspects. The technical working group is led by the Division of Applied Nuclear Physics at Uppsala University, where research on nuclear safeguards has been performed for over 30 years, and where competence in addition exists on a number of applied physics applications ranging from nuclear reactions, nuclear power and detection of radionuclides.

EURAD

The Differential Die-Away Self-Interrogation (DDSI) instrument detects neutrons in coincidence and is sensitive to the content of fissile material and neutron absorbing material in a fuel assembly. Its response to spent nuclear fuel with varying properties was investigated under the Next Generation Safeguards Initiative (NGSI) and the efforts included simulations as well as the manufacturing of a prototype instrument that was successfully tested in the field. Due to the interest of applying machine-learning techniques to support safeguards verification of spent nuclear fuel, we have investigated how to speed up the predictions of the early die-away time τ from the DDSI instrument for a large spent nuclear fuel inventory. One way to do this has been to develop a parametrisation function for τ as a function of the fuel parameters initial enrichment, burnup and cooling time. To assess the validity of the parameterisation function, the sensitivity of the DDSI response to various modelling parameters such as e.g. boron concentration, fuel pin geometry and operational history has been investigated. In this work, we summarise the recent efforts made to resolve these questions.

Under the Next Generation Safeguards Initiative, several different nuclear safeguards measurement techniques were studied. One of them was the Differential Die-AwaySelf-Interrogation technique, and the research showedthat its early die-away time τ was proportional to the fuel assembly multiplication and thus sensitive to the fissilecontent of the fuel assembly under assay. A prototype instrument was later built and tested in the field, and the measurements showed that the instrument could be usedsuccessfully in the field. This work builds on previous efforts, and systematically studies the effects of assumptions about the fuelproperties (such as its dimensions) and its irradiationconditions in the reactor, on the Rossi-Alpha Distribution(RAD) and τ. The motivation is twofold, firstly to betterunderstand if and what impacts such assumptions haveon the RAD and τ, and secondly to investigate how wellthe simulation model used to estimate the RAD and τ isable to generalize to other fuel types and irradiation conditions than those modelled. 20 spent nuclear fuel assemblies currently residing in theSwedish interim storage for spent nuclear fuel weremeasured by the prototype DDSI instrument. Theassemblies were modelled using Serpent2 and MCNP6 inthis work. Fuel depletion calculations were performed assuming both a standard irradiation cycle and the actualirradiation history as provided by the operator. Fuel properties and irradiation conditions were also modified and their effect studied. Based on the simulated DDSI instrument response inMCNP6, the RADs were created and τ determined. The analysis shows that each modelling assumption on itsown affects both the RAD and the τ value. However, some of the individual effects work in opposite direction and cancel out when considered at the same time. For this reason, the default model is considered to be a good and valid approximation of the more complex one and results are expected to generalize well.

In the current Digital Cherenkov Viewing Device (DCVD) measurement methodology, the DCVD is aligned over the centre of a fuel assembly when measuring emitted Cherenkov light. Due to the collimation of light, and due to the lifting handle of PWR fuel assemblies covering the fuel periphery, the DCVD is more sensitive to partial defects near the fuel assembly centre than near the periphery. Here, we investigate the sensitivity of the DCVD for detecting partial defects for different instrument alignments. By performing measurements at both the centre and near the assembly periphery, more accurate measurements near the periphery can be obtained. DCVD images were simulated for different partial defect scenarios with 30% of the fuel rods removed or replaced with low, medium or high-density rods. Simulations were run with different DCVD alignments, and the Cherenkov light distribution in the images were quantitatively analysed and compared to simulated images for a fuel assembly without defects. The simulation results were also compared with measurements of intact spent fuel assemblies. The simulations show that the local Cherenkov light intensity deviation due to a partial defect is not sensitive to the alignment. Hence, the current methodology is robust, and will not benefit from measuring at different alignments. Regarding the signal-to-noise ratio, combining measurements at different alignments can improve the measurements. However, the improvement is modest, and for the DCVD it may be preferred to simply use the current methodology and make longer measurements. For future autonomous Cherenkov measuring systems, combining images can be a way of improving the quality of the measurements.

The radionuclide composition of, and emitted radiation in, spent nuclear fuel from the future MYRRHA facility have been studied using depletion simulations to understand potential consequences for safeguards verification using non-destructive assay. The simulations show that both the gamma-ray and neutron emission rates in spent MYRRHA assemblies are lower than in spent PWR UO2 and MOX assemblies. In addition, gamma-ray emission rates from 134Cs and 154Eu are considerably lower, and the total neutron emission rate in MYRRHA fuel is much less sensitive to fuel burnup and cooling time. The main reason is that the fast neutron spectrum in MYRRHA affects the radionuclide production in the fuel. One result is that 244Cm, the main contributor to the neutron emission in spent light water reactor fuel, has a limited production in MYRRHA. Consequently, neutron-detection techniques could be used to more directly assay the plutonium content of spent MYRRHA fuel.

MYRRHA is an accelerator-driven system featuring a MOX-fuelled core cooled by lead-bismuth eutectic, which is under development at SCK CEN in Mol, Belgium. An initial plutonium content of 30% in the fuel is foreseen, which together with the fast neutron spectrum in the core results in considerably different spent-fuel properties compared to spent fuel from typical lightwater reactors. These differences have been studied through depletion simulations, and include how radionuclide densities depend on burnup, which radionuclides contribute to the gamma-ray and neutron emission, and the intensity of the emitted radiation. As a consequence, current techniques for safeguards verification of spent fuel via non-destructive assay may need to be updated, or new techniques developed, for use in safeguarding of spent MYRRHA fuel. To some extent, these differences may have consequences in the wider context of nuclear safeguards for Generation IV systems. The focus of this paper is to investigate to what extent the gamma-ray and neutron signatures could be used to differentiate irradiated MYRRHA fuel from irradiated MOX and UO2 fuels from a light water reactor. The ability to discriminate between light-water-reactor MOX and UO2 has been recognised as an important task in safeguards verification today, and this work extends this objective to a future nuclear energy system.

Nuclear safeguards verification of spent nuclear fuel (SNF) is imperative to ensure peaceful use of nuclear material. To verify the correctness and completeness of operator declarations, non-destructive assay (NDA) measurements of gamma and neutron radiation from the SNF play a central role. Verification of fuel based on such measurements is done routinely by safeguards inspectors and is also expected to be conducted prior to preparation of SNF for final disposal. Traditionally, SNF verification has been carried out by analyzing data from a single NDA instrument at a time. In this study, we compare the performances of different machine learning algorithms in their ability to make predictions of the SNF parameters such as fuel burnup (BU), initial enrichment (IE), and cooling time (CT). Predictions were made based on simulated signatures such as gamma-ray intensities from individual radionuclides, the total Cherenkov light intensity, and the parameterized differential die-away time (tau). In this work, multiple machine learning algorithms have been trained and tested on a set of simulated data containing 596,181 fuel samples providing a broad range of these three parameters to encompass the majority of the spent nuclear fuel worldwide. Additionally, the resilience of the machine learning algorithms on noisy data was evaluated. The results show that the non-linear methods can provide highly reliable predictions of SNF parameters. In nearly all situations assessed in this work, we have found that shallow learning methods have a clear advantage over deep learning models investigated in the present study. We also found that shallow learning methods such as k-Nearest Neighbors outperform other tree-based methods as well as neural networks at noise levels above 5%.

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

The paper describes a data library containing material composition of spent nuclear fuel. The data is extracted from burnup and depletion calculations with the Serpent2 code. The simulations were done with a PWR fuel pin cell geometry, for both initial UO2 and MOX fuel load for a wide range of initial enrichments (IE) or initial plutonium content (IPC), discharge burnup (BU) and cooling time (CT). The fuel library contains the atomic density of 279 nuclides (fission products and actinides), the total spontaneous fission rate, total photon emission rate, activity and decay heat at 789,406 different BU, CT, IE configurations for UO2 fuel and at 531,991 different BU, CT, IPC configurations for MOX fuel. The fuel library is organized in a publicly available comma separated value file, thus its further analysis is possible and simple.

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

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