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This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.
The Nuclear Material Management & Storage (MNN&S) and the Material Control and Accountability (MC&A) programs at the Nevada National Security Site (NNSS) have the responsibility to ensure accountable nuclear material (NM) no longer need by programs/projects is managed, coordinated, and dispositioned in accordance with DOE Orders 410.2, “Management of Nuclear Materials, and 474.2, “Nuclear Material Control and Accountability”. CD-NMM.007, “Programmatic Value Determination of Nuclear Material and Termination of Safeguards” is the document which lays out the process to fulfill these requirements. This paper defines the roles and responsibilities for all the organizations involved, the programmatic value determination (PVD), disposition planning, and termination of safeguards (ToS)
IAEA Facility-Level Safeguards and Implementation - This presentation is intended to introduce the audience to the application of international safeguards by the International Atomic Energy Agency at nuclear facilities around the world. It covers the State Level Approach, random statistical sampling processes, material balance area structure, non-destructive assay techniques, and inspection areas. Addition of advanced verification technologies for the IAEA
Increased purity in uranium ore concentrate not only raises the question as to whether Safeguards should be applied to the entirety of uranium conversion facilities, but also as to whether some degree of coverage should be moved back to uranium ore concentrate production at uranium mining and milling facilities. This paper looks at uranium ore concentrate production across the globe and explores the extent to which increased purity is evident and the underlying reasons. Potential issues this increase in purity raises for IAEA's strategy on the Starting Point of Safeguards are also discussed.
It is demanded to find key issues and critical paths for verifying and monitoring a spent nuclear fuel cask or storage with a purpose of non-proliferation of nuclear weapon. However, unfortunately there are no reliable methods to verify the containment in a spent nuclear fuel cask or storage. Therefore, research and development (R and D) has been continuously required to develop new and alternative technologies for searching and verifying undeclared nuclear materials, activities, and facilities. Therefore, the present authors have developed a new safeguards technique of passively scanning a spent fuel cask or storage using pressurized He-4 scintillation neutron detectors. A new approach is to quickly verify content in a spent nuclear fuel cask or presence of all special nuclear materials in unknown packages by comparing experimental and computational results. For that, the present authors have developed a new algorithm and user-interface coupled ORIGEN-ARP and MCNP codes, and the prototype safeguards apparatus employing the array of He-4 fast neutron detectors has been designed, built, and tested. The feasibility test of this new apparatus has been performed with Cf-252 neutron sources and the laboratory scale model cask, minimized by a factor of ten compared to the actual dimension of a KN-21 cask. This technique has several advantages over current methods based on neutron and gamma detection; (1) incident neutron energy information is preserved, (2) disturbance of gamma-ray is excellently rejected (discriminated), and (3) much higher neutron counts and operational cost and space-effectiveness are expected since no moderator is required for fast neutrons. The final goal of this study will (1) conduct a prototype instrument and employ it to monitor a nuclear spent fuel cask or storage at a commercial nuclear power reactor facilities, (2) use spectral-specific measurement data to show proof of concept for measurable data successfully converted to quantifiable signatures, such as diversion/removal of fuel, and maintenance of fuel and cask integrity of fuel, and (3) develop a cask imaging capability. (authors)
Pyroprocessing presents several new challenges for materials accountability and the overall design of a safeguards approach. While measurement techniques for aqueous plants have been optimized over several decades, the material and process differences for pyroprocessing require new approaches. This paper summarizes the safeguards challenges and a preliminary approach for materials accountability. (authors)
Safeguarding spent fuel assemblies is one of the most important objectives for nuclear non-proliferation, since it contains fissile material which can be diverted into nuclear explosive devices. Partial defect detection is a challenging issue among IAEA future safeguards requirements. The IAEA defines there exist partial defects if 50% or more of the fuel pins in a spent fuel assembly are diverted. Conventional partial defect detectors are classified into two groups depending on its operating principle. The first group applies gamma/neutron spectrometer, which are PDET, SMOPY, and gamma emission tomography. Partial defect detectors in the first group are able to calculate the burnup, isotope composition, and cooling time of a spent fuel assembly. However, since it takes time to analyze a spent fuel assembly, they cannot detect every spent fuel assemblies in a cooling pool. The second group applies Cerenkov radiation detector, which can screen a spent fuel assembly within a short time. However, since Cerenkov radiation cannot be generated under atmospheric environment, it cannot be applied for dry condition. This research proposes a scintillator based partial defect detector (SPDD) for inspecting spent fuel, which can detect partial defects within a short time and both in and out of a cooling pool.
The proposed poster focuses on using acoustic monitoring to advance detection techniques for reprocessing equipment in support of nuclear safeguards. The usage of free air acoustic monitoring has been previously demonstrated at Idaho National Laboratory (INL) at facilities such as the Advanced Test Reactor and the National Security Test Range. However, this work focuses on a deployment environment dedicated to monitoring separation processes, using solvent extraction equipment such as centrifugal contactors. This environment offers an opportunity for signal discovery and in characterizing acoustic signatures of the equipment in operation. This data can support the development of safeguards by design and security by design measures for aqueous reprocessing facilities. Furthermore, this type of monitoring can aid in early detection and identification of removed materials indicating diversion, which is essential for initiating material recovery and actor identification. The results of this work include data from nine low-frequency acoustic sensors used to investigate potential acoustic characteristics of centrifugal contactors used in multi-stage processes. The results also discuss the correlation between the signatures and the operation of the contactor banks.
The application of a triple bubbler system within molten salt environments, specifically a eutectic mixture of LiCl-KCl, is explored for nuclear safeguards purposes, focusing on nuclear material accountancy (NMA) and holdup accountancy in pyroprocessing techniques. Traditional holdup accountancy methods, which involve periodic sampling and lengthy lead times, are being supplemented by the development of actinide sensors using electrochemistry. However, these methods alone are insufficient to meet regulatory requirements set by agencies like NRC and IAEA. The triple bubbler device measures salt level, density, and surface tension simultaneously using three dip tubes of varying diameters and heights. This study investigates the accuracy and robustness of the triple bubbler in molten salt conditions at temperatures ranging from 450-550°C. The updated bubbler design, utilizing tantalum metal and larger tube geometry, demonstrated minimal plugging and maintained measurement accuracy within acceptable uncertainties: 0.611% for depth, 1.34% for density, and approximately 40% for surface tension. Results indicate that the triple bubbler system is effective for real-time, accurate measurements in pyroprocessing environments, making it a promising tool for ensuring compliance with nuclear safeguards and supporting nuclear non-proliferation efforts.
Recent developments in nuclear fuel reprocessing techniques have yielded more efficient processes and fuel cycle options that strengthen the nuclear industry and production of clean energy. One such area of interest is pyroprocessing of used oxide fuel. However, with these advances in the back end of the nuclear fuel cycle, advances in safeguards instrumentation, measurements, and approaches are needed to ensure special nuclear material (SNM) is accounted for according to regulatory requirements. As a high-level overview of a nominal pyroprocessing approach, used oxide fuel from commercial light water reactors (LWR) is mechanically removed from the metallic cladding. Then the fuel is crushed and randomized representative samples are taken and sent to an analytical lab for analysis. The analytical results of the feed material are used for input accountancy into the rest of the process. The crushed oxide fuel is then moved to the oxide reduction (OR) furnace where it is reduced from an oxide to metallic form. The OR product is distilled to remove salt and then moved to an electrorefiner (ER), where it is immersed in a eutectic mixture of lithium chloride potassium chloride (LiCl-KCl) that typically ranges between 450-550 ?. Within the ER, the usable uranium is electrochemically transported through the molten salt from the anode to the cathode, and then subsequently removed as a relatively pure U product.. A simplified model of pyroprocessing techniques with added emphasis on the safeguards can be seen below in Fig. 1
Here, we extend the familiar Bӧhnel point-model equations, which are routinely used to interpret neutron coincidence counting rates, by including the contribution of delayed neutrons. After developing the necessary equations we use them to show, by providing some numerical results, what the quantitative impact of neglecting delayed neutrons is across the full range of practical nuclear safeguards applications. The influence of delayed neutrons is predicted to be small for the types of deeply sub-critical assay problems which concern the nuclear safeguards community, smaller than uncertainties arising from other factors. This is most clearly demonstrated by considering the change in the effective (α,n)-to-spontaneous fission prompt-neutron ratio that the inclusion of delayed neutrons gives rise to. That the influence of delayed neutrons is small is fortunate, and our results justify the long standing practice of simply neglecting them in the analysis of field measurements.
Multidimensional gamma-spectrometry is a potential solution for safeguarding future thorium-fueled molten salt reactors (MSRs). A high-sensitivity system developed at Pacific Northwest National Laboratory (PNNL), the Advanced Radionuclide Gamma spectrOmeter (ARGO), has demonstrated direct 233 U assay in freshly irradiated fuel salt samples using a dual-detector configuration and the 97.1 keV gamma-emission. These measurements were made using 1.4 x 10 6 Bq of 233 U activity (at 1.4 x 10 8 fissions), at relatively high dead times (10.4%), and using short count durations of ~800 s. Other major emissions were also evaluated (42.5 keV, 54.7 keV, 146.4 keV), in additional to limited (1.96%) Compton suppression, but provided less consistent determination of the 233 U activity. Accurate 233 U measurement was also demonstrated using its four coincidence gamma-emissions (42.5 keV, 54.7 keV; 42.5 keV, 278.1 keV; 146.4 keV, 174.2 keV; 146.4 keV, 219 keV). Further improvements in coincidence measurements may be achievable with improvements in the detector configuration and nuclear data. Whilst this demonstration utilized the ARGO laboratory system, a similar dual-detector configuration could be utilized for online safeguards measurements.
Californium-252 is an isotope commonly encapsulated as a physically small but strong spontaneous fission neutron source for applications in industry, academia, and research. Within the nuclear safeguards community, a well-characterized 252 Cf source is often used as an accessible alternative for plutonium in calibrations measurements, which require the absolute source strength. Current methods of source strength quantification can provide an accurate estimate, approximately 1% uncertainty under optimal conditions, but are typically costly and time consuming. An alternative method was developed to determine 252 Cf source strength using passive neutron correlation counting (PNCC). The PNCC method enables institutions and facilities to perform 252 Cf source calibration measurements in-house using detection systems that are common within the nuclear safeguards community. This novel method was previously successfully demonstrated using neutron multiplicity well counters. However, the previous evaluation utilized detection systems with high neutron detection efficiencies, greater than 60%. The purpose of the current study is to extend the previous evaluation and demonstrate the feasibility of the PNCC 252 Cf calibration method for suite of coincidence counters with lower neutron detection efficiencies, between 15% and 35%, which are more commonly encountered in the field, as well as to provide a step-by-step guide to using the method. The neutron source used in this evaluation was previously certified by the National Institutes of Science and Technology (NIST), which provides a reference for the accuracy of the PNCC method. In conclusion, the present source yield calculations demonstrate excellent agreement with a NIST calibration technique and achieve a standard uncertainty below the target 1%.
Nuclear fuel reprocessing plants contain a wealth of plant monitoring data including material measurements, process monitoring, administrative procedures, and physical protection elements. Future facilities are moving in the direction of highly-integrated plant monitoring systems that make efficient use of the plant data to improve monitoring and reduce costs. The Separations and Safeguards Performance Model (SSPM) is an analysis tool that is used for modeling advanced monitoring systems and to determine system response under diversion scenarios. This report both describes the architecture for such a future monitoring system and present results under various diversion scenarios. Improvements made in the past year include the development of statistical tests for detecting material loss, the integration of material balance alarms to improve physical protection, and the integration of administrative procedures. The SSPM has been used to demonstrate how advanced instrumentation (as developed in the Material Protection, Accounting, and Control Technologies campaign) can benefit the overall safeguards system as well as how all instrumentation is tied into the physical protection system. This concept has the potential to greatly improve the probability of detection for both abrupt and protracted diversion of nuclear material.
During the past several years, the demand for 3 He gas has far exceeded the gas supply. This shortage of 3 He gas is projected to continue into the foreseeable future. To continue to develop and deploy safeguards neutron detection systems, a viable, near-term, safeguards-specific, alternative to 3 He-based measurement technology must be found.
This report summarizes the 2014 Next Generation Safeguards Professional Network (NGSPN) meeting held at Sandia and Los Alamos National Laboratories. The meeting comprised two days at each laboratory and included facility tours, technology demonstrations, presentations, and networking opportunities to strengthen expertise in nuclear safeguards and nonproliferation.
GCEPS safeguards (Desire to close gaps on undeclared feed) - Timely detection of the misuse of the facility to produce HEU - Timely detection of the diversion of declared UF6 - Timely detection of the misuse of the facility to produce undeclared LEU – Operator “no one would ever divert undeclared LEU product” – Breakout vs. Clandestine Ops (CSA and AP safeguards options) • Tightened coverage of GCEPs – declared operations (SNRI,C/S) • Unattended monitoring system – in development for GCEPs – Load Cells – Monitoring cylinders / weight – On-Line Enrichment Monitoring – Cylinder Verification Station – NDA, Weight,… – UF6 Cylinder Tracking - UID • Sensitive technology! Khan network and undeclared facilities
Many nuclear safeguards applications could benefit from high-resolution gamma-ray spectroscopy achievable with metallic magnetic calorimeters. This dissertation covers the development of a system for these applications based on gamma-ray detectors developed at the University of Heidelberg. It demonstrates new calorimeters of this type, which achieved an energy resolution of 45.5 eV full-width at half-maximum at 59.54 keV, roughly ten times better than current state of the art high purity germanium detectors. This is the best energy resolution achieved with a gamma-ray metallic magnetic calorimeter at this energy to date. In addition to demonstrating a new benchmark in energy resolution, an experimental system for measuring samples with metallic magnetic calorimeters was constructed at Lawrence Livermore National Laboratory. This system achieved an energy resolution of 91.3 eV full-width at half-maximum at 59.54 keV under optimal conditions. Using this system it was possible to characterize the linearity of the response, the count-rate limitations, and the energy resolution as a function of temperature of the new calorimeter. With this characterization it was determined that it would be feasible to measure 242Pu in a mixed isotope plutonium sample. A measurement of a mixed isotope plutonium sample was performed over the course of 12 days with a single two-pixel metallic magnetic calorimeter. The relative concentration of 242Pu in comparison to other plutonium isotopes was determined by direct measurement to less than half a percent accuracy. This is comparable with the accuracy of the best-case scenario using traditional indirect methods. The ability to directly measure the relative concentration of 242Pu in a sample could enable more accurate accounting and detection of indications of undeclared activities in nuclear safeguards, a better constraint on source material in forensic samples containing plutonium, and improvements in verification in a future plutonium disposition treaty.