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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Remediation of the FGMSP at Sellafield: The First Transfer of Bulk Sludge - 17133

The very first bulk transfers of radioactive sludge have commenced from one of the most hazardous nuclear plants in Europe, using equipment and systems researched, designed, installed and commissioned by Jacobs as part of the ACKtiv alliance. Sellafield is a nuclear fuel reprocessing and decommissioning site, close to the village of Seascale on the coast of the Irish Sea in Cumbria, England. Sellafield incorporates the original nuclear reactor site at Windscale, which is currently undergoing decommissioning, and dismantling, and Calder Hall, the world's first nuclear power generating reactor. The first nuclear facilities were built in the 1940's.The First Generation Magnox Storage Pond (FGMSP) dates back to the 1950's as part of the UK's expanding nuclear programme and was constructed to store, cool and prepare used Magnox nuclear fuel for recycling into new fuel. It urgently needs to be emptied of 1500 cubic metres of radioactive sludge. In this paper we describe the work undertaken by Jacobs to deliver a very challenging programme of work to provide the capability to remove sludge from the pond. The FGMSP Retrievals Project was awarded in March 2007 to supply equipment and services for the removal of sludge and miscellaneous waste from the First Generation Magnox Storage Pond at Sellafield. The Project is being executed by ACKtiv Nuclear (Jacobs, Carillion and Atkins) of which Jacobs is a 40% stakeholder. ACKtiv Nuclear is a mature Joint Venture and has been operating continuously and successfully since 2002, the oldest JV still in operation on Sellafield Site. The roots of the JV can be traced back to the 1980's. The Project covers all project phases from front-end engineering through detailed design, procurement, installation and commissioning support. The Project places huge emphasis on off-Site testing and has several facilities for both installation and integrated testing. This proves equipment prior to installation in a radioactive environment. The Project has achieved success and recognition including winning the 2014 and 2015 Sellafield Business Excellence awards, as well as numerous internal Jacobs awards. The Project achieved a major decommissioning milestone in March 2016. Since the installation of a 31 m pipe-bridge in 2012, using a 1200 te mobile crane surrounded by several other facilities with nuclear inventory, the Project has successfully installed the sludge retrieval and effluent system into the FGMSP. This 4 year programme of work included 2 booster pump platforms, sludge recovery tooling, hoses, umbilicals and a deluge skip wash box which were all trialled and proven off-Site prior to installation in the pond. This milestone to commence transfer of sludge to modern stainless steel containment is a major step forward in the hazard reduction of the facility.The presentation will cover the installation of the pond de-sludging equipment, the innovations deployed that ensured that the milestone was achieved under budget and ahead of schedule, as well the benefits to the facility, the Site and the nuclear industry. It will also examine the challenges that needed to be overcome such as high dose, contamination, limited working times and space constraints. The presentation will describe why the work is of national Importance and conclude with the achievement of commencement of bulk sludge transfer to a modern facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Human Factors and Nuclear Criticality Safety Accidents [Slides]

The agenda for the presentation is as follows: Human Factors; Accident Review (Tokai Mura, Siberian Chemical Combine); Procedures; Accident Review (Los Alamos Waste Recovery); Environmental Distractions; Accident Review (Idaho Fuel Reprocessing); and, Equipment Reliability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Co-adsorption of I{sub 2} and H{sub 2}O on Ag{sup 0}Z and Kinetics of Iodine Silver Reaction

Co-adsorption of iodine and water by reduced silver-exchanged mordenite (Ag{sup 0}Z) is studied to support the development of multi-component adsorption models for the separation and recovery of radio isotopic gases in off-gas streams in nuclear fuel reprocessing. Single-component adsorption equilibrium data of iodine and water were obtained to determine the saturation capacity and evaluate the parameters of the generalized statistical thermodynamic adsorption (GSTA) model being developed for modeling of co-adsorption. Co-adsorption data acquisition is in progress for model validation. Kinetic data were obtained to determine the intra-particle mass transfer and reaction parameters for the adsorption models. In addition, work is in progress to obtain data of the iodine-silver reaction at various conditions to study the reaction kinetics, which will provide insight to the chemisorption process of iodine on Ag{sup 0}Z. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Removal of Tritium from the Gaseous Effluents Produced by the Recovery of Zirconium from Used Fuel Cladding

U.S. regulations will require the removal of tritium ({sup 3}H) from gaseous effluents associated with nuclear fuel reprocessing and recycling. One process that may generate {sup 3}H-containing gaseous effluents is the recovery of zirconium from used fuel cladding. A chlorination process developed by Collins et al. is designed to convert the zirconium present in Zircaloy{sup R} cladding to recoverable ZrCl{sub 4}. The greatest economic benefit from including zirconium recycle as part of a larger used nuclear fuel (UNF) recycle effort would be a decrease in the disposal cost of used fuel cladding. Used fuel cladding is considered high-level waste (HLW). As zirconium is the second largest element by mass (25%) contained in most UNF assemblies (with uranium being the largest), recovery of the zirconium could result in a substantial decrease in the amount of HLW intended for geologic disposition. The reclaimed ZrCl{sub 4} product, if sufficiently free from radioactive impurities, could then be disposed of as low-level waste or reused in nuclear applications. The chlorination process uses chlorine gas (Cl{sub 2}) and an inert carrier such as argon (Ar) to convert the zirconium present in cladding to ZrCl{sub 4}. ZrCl{sub 4} is more volatile than other components in the cladding, causing evaporation of a relatively pure gaseous ZrCl{sub 4} stream. The product ZrCl{sub 4} is recovered as a solid salt by condensation when cooled to less than about 300 deg. C. A fraction of the {sup 3}H (0-96%) produced in nuclear fuel during irradiation is found in zirconium-based cladding, possibly as zirconium hydride (the wide range is a remnant of various researchers' analyses of cladding of various irradiations and cooling times). Heating the cladding to high temperatures, such as would be done during the chlorination process, will release the {sup 3}H. At the temperatures of the zirconium recycle process, the {sup 3}H is likely to react with excess Cl{sub 2} to produce tritiated HCl ({sup 3}HCl), although this could be avoided if the cladding and fuel were treated to remove {sup 3}H prior to chlorination. A previous analysis showed that the concentration of {sup 3}HCl in the off-gas of the zirconium recycle process will be in the range of parts per billion to a few parts per million. Should {sup 3}H mitigation be required for the zirconium recycle process, it will be completed through the abatement of {sup 3}HCl from this off-gas stream. The intent of the testing described in this summary was to conduct proof-of-principle demonstrations for two {sup 3}HCl removal methods. The zirconium recycle off-gas stream differs from many industrial HCl-containing streams in that the concentration of HCl requiring removal is extremely low. Demonstrations of HCl removal at ppm concentrations were not found in the literature, making the proof-of-principle approach attractive as a way to identify the most promising method for continued development. An important consideration throughout these experiments was that current research efforts on the zirconium recycle process are conducted in a hot cell and any {sup 3}HCl removal methods should be easily deployed in a hot-cell environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Decommissioning Experience at Rokkasho; Standing Type Manipulator: A1000S for Dismantling Tasks - 16434

Decommissioning is a technical process which includes various steps and processes from the clean-up of radioactive materials to the progressive demolition of the plant. One of these steps is the cutting of subassemblies, like tanks, pipes or walls in small pieces suitable for the sorting and disposal. Different solutions are offered for this cutting work and can be grouped as following: - Chip forming - sawing, drilling, milling, turning etc. - Shearing - punching, stamping, scissoring. - Abrading - Grinding, lapping, polishing; water-jet. - Heat - flame cutting, plasma cutting, laser cutting. - Electro-Chemical - Etching, Electrical discharge machining (EDM). Each method has its limitations in accuracy, cost, and effect on the material. In 2013, Waelischmiller Engineering won a contract for two remote handling systems dedicated to the dismantling cell at Rokkasho Nuclear Fuel Reprocessing Facility in Japan. For this project, Inconel and other metals needed to be cut. One of the main cutting methods was decided to use of an angle grinder handled by a robotic arm. Because the use of the overhead systems (bridges and cranes) was difficult for this installation, it was determined, that the best alternative was a power manipulator A1000 installed on the floor with a pre-positioning system. The project proceeded to go forward with the adaptation of the standard A1000 with integration of a control system in a Cartesian coordinate frame; resulting in a heavy duty arm; A1000S with 6 degrees of freedom. The heavy duty power arm A1000S was developed and designed for mechanical cutting with an angle grinder, especially for thick and hard material for example stainless steel and Inconel. Cutting a thick material with an angle grinder requires repeatable precise linear motion, in order not to damage the disc. The angle grinder should be handled with a manipulator which can cut in all directions in a linear motion, since the objects which should be cut are usually very large and fixed in any direction For these requirements, the Cartesian control function was added to our standard power manipulator type A1000. In some cases, the decommissioning work includes repeated access routes or fixed access points that the manipulator arm has to pass repeatedly. For example a fixed point where the waste materials are collected, or repeated transport route that the arm has to go along for each step. For this requirement, the A1000S has been equipped with the 'Go To-Mode'. It is a 'teach and playback mode' which has been developed especially for remote operating manipulators. Quick direct teach, quick play function and an intuitive touch-panel layout are some of the features. With the Go To-Mode high working efficiency was achieved. After rigorous testing at Waelischmiller premises; the manipulator was delivered to JNFL in Rokkasho. It was found that the obtained functionality and precision can be applicable not only for the cutting with a disc grinder, but also the laser cutting method, which requires precise trajectory control. This paper will provide: - A project description, - Presentation of solution process, - Technical description of the A1000S, - Future Applications of the robotic Arm A1000S. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Ruthenium Volatilisation in Nuclear Waste Systems -Comparison of Ru(III)Cl{sub 3} and Ru(III)NO(NO{sub 3}){sub 3} - 17113

Ruthenium is a fission product possessed of two relatively long lived isotopes; {sup 103}Ru and {sup 106}Ru, both of which form part of the Highly Active (HA) waste raffinate during spent nuclear fuel reprocessing. During reprocessing, ruthenium, which may be in the form of the RuNO{sub 3}+ complex, encounters temperatures conducive to volatilisation. Due to ruthenium's high specific radioactivity it is important to understand the mechanism by which volatilisation occurs. Here we use combined cyclic voltammetry (CV) and electrochemical micro-gravimetry experiments in a study of both NO free Ru (III), in the form of a RuCl{sub 3} solution and NO complexed Ru (III), in the form of RuNO{sub 3}+ solutions. We do this in order to establish baseline behaviour of NO free Ru (III) and compare it to NO complexed Ru (III) to determine the effect of complexation on the system. We have observed discrete oxidations in both systems from solution phase Ru (III) to solid phase RuO{sub 2} and volatile RuO{sub 4}. Two key differences emerged between the two systems, the first being the mass changes associated with deposition and stripping of RuO{sub 2} are significantly smaller in the RuNO{sub 3}+ solutions than those for solutions of RuCl{sub 3}. The second difference is the difference in onset of volatile RuO{sub 4} generation, in the RuNO{sub 3}+ solution, RuO{sub 4} generation is 100 mV positive than that for the RuCl{sub 3} solution. This suggests NO complexation protects Ru (III) against oxidation and, to a certain extent, volatilisation as RuO{sub 4}. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Direct measurement of 235 U in spent fuel rods with Gamma-ray mirrors

We report here that direct measurement of plutonium and uranium X-rays and gamma-rays is a highly desirable nondestructive analysis method for the use in reprocessing fuel environments. The high background and intense radiation from spent fuel make direct measurements difficult to implement since the relatively low activity of uranium and plutonium is masked by the high activity from fission products. To overcome this problem, we make use of a grazing incidence optic to selectively reflect K α and K β fluorescence of Special Nuclear Materials (SNM) into a high-purity position-sensitive germanium detector and obtain their relative ratios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Review of recent developments in iodine wasteform production

Radioiodine capture and immobilization is not only important to consider during the operation of reactors (i.e., I-131), during nuclear accidents (i.e., I-131 and I-129) or nuclear fuel reprocessing (i.e., I-131 and I-129), but also during disposal of nuclear wastes (i.e., I-129). Most disposal plans for I-129-containing waste forms (including spent nuclear fuel) propose to store them in underground repositories. Here, iodine can be highly mobile and, given its radiotoxicity, needs to be carefully managed to minimize long-term environmental impacts arising from disposal. Typically, any process that has been used to capture iodine from reprocessing or in a reactor is not suitable for direct disposal, rather conversion into a wasteform for disposal is required. The objectives of these materials are to use either chemical immobilization or physical encapsulation to reduce the leaching of iodine by groundwaters. Some of the more recent ideas have been to design capture materials that better align with disposal concepts, making the industrial processing requirements easier. Research on iodine capture materials and wasteforms has been extensive. This review will act as both an update on the state of the research since the last time it was comprehensively summarized, and an evaluation of the industrial techniques required to create the proposed iodine wasteforms in terms of resulting material chemistry and applicability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Determination of plutonium in spent nuclear fuel using high resolution X-ray

Characterization of Pu is an essential aspect of safeguards operations at nuclear fuel reprocessing facilities. A novel analysis technique called hiRX (high resolution X-ray) has been developed for the direct measurement of Pu in spent nuclear fuel dissolver solutions. hiRX is based on monochromatic wavelength dispersive X-ray fluorescence (MWDXRF), which provides enhanced sensitivity and specificity compared with conventional XRF techniques. A breadboard setup of the hiRX instrument was calibrated using spiked surrogate spent fuel (SSF) standards prepared as dried residues. Samples of actual spent fuel were utilized to evaluate the performance of the hiRX. The direct detection of just 39 ng of Pu is demonstrated. Initial quantitative results, with error of 4–27% and precision of 2% relative standard deviation (RSD), were obtained for spent fuel samples. The limit of detection for Pu (100 s) within an excitation spot of 200 μm diameter was 375 pg. This study demonstrates the potential for the hiRX technique to be utilized for the rapid, accurate, and precise determination of Pu. Moreover, the results highlight the analytical capability of hiRX for other applications requiring sensitive and selective nondestructive analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Aging of Iodine-Loaded Silver Mordenite in NO2

Used nuclear fuel facilities need to control and minimize radioactive emissions. Off-gas systems are designed to remove radioactive contaminants, such as 85Kr, 14C, 3H, and 129I. In an off-gas system, any capture material will be exposed to a gas stream for months at a time. This gas stream may be at elevated temperature and could contain water, NOx gas, or a variety of other constituents comprising the dissolver off-gas stream in a nuclear fuel reprocessing plant. For this reason, it is important to evaluate the effects of long-term exposure, or aging, on proposed capture materials. One material under consideration is reduced silver mordenite (Ag0Z), which is recognized for its efficient iodine capture properties. Iodine is immobilized on Ag0Z as AgI, a solid with low volatility (m.p. ≥ 500°C). The aim of this study was to determine whether extended aging at elevated temperature in a nominally 2% NO2 environment would result in a loss of immobilized iodine from this material due to either physical or chemical changes that might occur during aging. Charges of iodine-loaded reduced silver mordenite (I2-Ag0Z) were exposed to a 2% NO2 environment for 1, 2, 3, and 4 months at 150°C, then analyzed for iodine losses The aging study was completed successfully. The material did not visibly change color or form. The results demonstrate that no significant iodine loss was observed over the course of 4 months of 2% NO2 aging of I2-Ag0Z at elevated temperature within the margin of error and the variability (~10%) in the loading along the beds. This provides assurance that iodine will remain immobilized on Ag0Z during extended online use in an off-gas capture treatment system. Future tests should expose I2-Ag0Z to progressively more complex feed gases in an effort to accurately replicate the conditions expected in a reprocessing facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Recycling facility periodic safety review: Application of a methodology used for the nuclear criticality safety review of geometrically safe or favorable equipment - 14610

In France, for all fuel cycle facilities, the reference safety frame is continuously adapted to take into account technical developments such as changes in process and/or maintenance operations. Every ten years according to French law, the facility operator is required to launch a process of Periodic Safety Review (PSR), whose conclusions are sent to the Nuclear Safety Authority (NSA). This PSR is split in two parts: first, a review of the facility to its safety documentation and authorization, second, a safety reassessment. Actually, this PSR takes account for topics (actual aging for instance) which require several years of feedback from experience to be integrated in safety evaluations, or which cannot be done on a daily basis (documentation reorganization). It also incorporates the new regulatory safety requirements. A specific methodology was developed for the first PSR of UP3-A since the publication of the law. UP3-A is a used fuel reprocessing facility of La Hague plant. This methodology, adapted for large plants with numerous process and fissile media, is based on a two-step approach intending to focus the analysis on the real hard spots. First, 'sensitive' systems are identified. Second, a specific review is performed on these systems. After a reminder of the methodology, this paper presents the results of its application on the UP3-A scope. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Spent Nuclear Fuel Disposition

One interdisciplinary field devoted to achieving the end-state of used nuclear fuel (UNF) through reuse and/or permanent disposal. The reuse option aims to make use of the remaining energy content in UNF and reduce the amount of long-lived radioactive materials that require permanent disposal. The planned approach in the U.S., as well as in many other countries worldwide, is direct permanent disposal in a deep geologic repository. Used nuclear fuel is fuel that has been irradiated in a nuclear reactor to the point where it is no longer capable of sustaining operational objectives. The vast majority (by mass) of UNF is from electricity generation in commercial nuclear power reactors. Furthermore, the other main source of UNF in the U.S. is the Department of Energy’s (DOE) and other federal agencies’ operation of reactors in support of federal government missions, such as materials production, nuclear propulsion, research, testing, and training. Upon discharge from a reactor, UNF emits considerable heat from radioactive decay. Some period of active on-site cooling (e.g., 2 or more years) is typically required to facilitate efficient packaging and transportation to a disposition facility. Hence, the field of UNF disposition broadly includes storage, transportation and ultimate disposition. See also: Nuclear Fission (content/nuclear-fission/458400), Nuclear Fuels (/content/nuclear-fuels/458600), Nuclear Fuel Cycle (/content/nuclear-fuel-cycle/458500), Nuclear Fuels Reprocessing (/content/nuclear-fuels-reprocessing/458700), Nuclear Power (/content/nuclear-power/459600), Nuclear Reactor (/content/nuclear-reactor/460100), Radiation (/content/radiation/566300), and Radioactive Waste Management (/content/radioactive-waste-management/568900).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preparation of Waste Fingerprints for the Miscellaneous Beta Gamma Waste Feeds to the Box Encapsulation Plant at Sellafield - 16080

There are a number of legacy pond and silo storage facilities on Sellafield Site that contain a broad range of ILW arising from the early Windscale Pile and Magnox Fuel Reprocessing Programmes. The waste comprises of solid Miscellaneous Beta Gamma Waste (MBGW) from operations (including filters, containerised waste, thermocouple wire, support struts, pumps, soft wastes) and associated sludges that have arisen from the corrosion of the fuel, Magnox Swarf cladding and the MBGW. It is intended to retrieve and package this waste to meet the requirements of interim storage and final disposal in the Box Encapsulation Plant (BEP) at Sellafield. As part of this process each package must be assigned a realistic and justifiable radionuclide and material inventory. These inventories can either be assigned by radiometric measurements including the associated errors or developed by fundamental data underpinning these assessments and a 'chain of custody' approach of knowing where the waste has been and how it has evolved. The subject of this paper is concerned with the development of the waste fingerprints for these legacy waste streams. As the wastes have arisen over a ∼70 year period there have been an assortment of different working practices and developments in equipment design that have led to significant variations in the waste condition and its composition. During storage the waste items have become mixed, corroded and sludge deposits have arisen. Because of this, and that early consignment records are either incomplete or missing, the task of assigning material and radionuclide fingerprints has been very challenging. In addition, to comply with regulatory requirements the inventory assignment has to be sufficiently robust to meet the processing envelope of the encapsulation process i.e. to allow controls to be made on the amounts of reactive materials (e.g. aluminium, Magnox and uranium) that can be processed within each product and allow identification of excluded items from being consigned from the donor plants. Also, as the waste comprises actinides, fission and activation products it was noted that whilst the UK standard inventory code FISPIN10 predicts the actinides and fission products well, there was a significant discrepancy in the prediction of activation products when compared with those predicted by the specialised fusion code FISPACT. It should be noted that the use of FISPACT to predict actinides and fission products is not recommended. To address these issues, - position statements were produced for each of the waste streams that would be consigned to BEP; - gaps within the data were assessed and where possible inactive trial work was identified to address it; - material and radionuclide fingerprint justification documents were developed from the position statements based upon available data, the experimental work and justified assumptions; - where FISPIN runs were required, these were run with a modified activation data library based on the European Activation File 2010 (EAF2010) to model the activation products. This paper summarises the BEP process, discusses examples of how the material and radionuclide waste stream fingerprints have been justified by revisiting the sample data, original consignment records, engineering drawings and recent remotely operated vehicle (ROV) pond surveys and making reasoned assumptions based on operational practices and non-active experimental work. Incorporation of the EAF2010 library within the FISPIN code has offered significant benefits to BEP and indeed other plants processing mixed fission and activated wastes as it allows comparable predictions of activation products compared with that made by FISPACT and for the plant to maintain one data processing stream rather than two if it had to utilise both modelling codes. This work will underpin the inventory submission in support of the Interim Letter of Compliance (ILoC) to Radioactive Waste Management Limited (RWM). The process of examining the original records, plant operation schedules and engineering drawings has enabled removal of some of the pessimisms that have been made in previous assessments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Advanced Nanomaterials for Nuclear Energy and Nanotechnology

The use of the latest engineered nanomaterials in nuclear energy systems has opened doors for improving the performance and safety of nuclear power. Nuclear nanotechnology (NNT) deals with the use of engineered nanomaterials for future nuclear energy applications. In this paper, the recent progress in research and benefits of using nanomaterials in different areas of nuclear energy like nuclear fuel extraction and fabrication, fission product capturing, creating robust reactor materials, radiation sensing and monitoring, and radioactive waste separation and spent nuclear fuel reprocessing are summarized. Ongoing research around the world in this field is reviewed, including experimental and computational methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Statistical Approaches for Pebble Bed Reactor Operations and Safeguards

The design of pebble bed reactors (PBRs) and their method of operation align more closely with statistical approaches used in manufacturing and process control than traditional safeguards statistical approaches. The reason is PBRs will employ a nondestructive assay (NDA) measurement (burnup measurement system [BUMS]) that is part of the fuel handling system supporting discharge decisions in addition to the reactor code which monitor performance. The integration of these two approaches provides the opportunity to monitor reactor performance statistically for both operations and safeguards in ways not achievable using other reactor designs. For light water reactors (LWRs), knowledge about reactor code performance in predicting irradiated special nuclear material (SNM) content historically was only achieved from special measurement campaigns or from fuel reprocessing. Conversely, through statistical comparison of the BUMS with the reactor code-predicted values, PBRs can achieve this in real time. The resulting SNM distribution is an indicator of reactor performance because factors such as transit time and path of the pebble fuel through the reactor determine the plutonium production and uranium depletion. By analyzing the predicted and measured values, opportunities exist to adjust operating parameters, fuel design, and other characteristics to optimize performance and fuel utilization. From a safeguards perspective, this approach also provides the information necessary to validate declared values and evaluate whether the reactor is being operated as expected. This paper outlines statistical approaches for PBRs that can be used to support both operations and safeguards.

Ball, Cory↗