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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Experimental and modeling studies of PR and ND oxalate solubility to high ionic strengths: Insight into actinide(III) oxalates

Actinide oxalates are chemical compounds important to nuclear industry, ranging from actinide separation in waste reprocessing, to production of specialty actinides, and to disposal of high level nuclear waste (HLW) and spent nuclear fuel (SNF). In this study, the solubility constants for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O by performing solubility experiments in HNO 3 and mixtures of HNO 3 and H 2 C 2 O 4 at 23.0 ± 0.2 °C have been determined. The targeted starting materials, Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O, were successfully synthesized at room temperature using PrCl 3 , NdCl 3 and oxalic acid as the source metrials. Then, we utilized the targeted solubility-controlling phases to conduct solubility measurements. There was no phase change over the entire periods of experiments, demonstrating that Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O were the solubility-controlling phases in our respective experiments. Based on our experimental data, we have developed a thermodynamic model for Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O in the mixtures of HNO 3 and H 2 C 2 O 4 to high ionic strengths. The model for Pr 2 (C 2 O 4 ) 3 •10H 2 O reproduces well the reported experimental data for Pu 2 (C 2 O 4 ) 3 •10H 2 O, which are not utilized for the model development, demonstrating that Pr(III) is an excellent analog for Pu(III). Similarly, the model for Nd 2 (C 2 O 4 ) 3 •10H 2 O reproduces the solubility of Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The Pitzer model was used for the calculation of activity coefficients. Based on the published, well established model for dissociation constants for oxalic acid and stability constants for actinide-oxalate complexes [i.e., AmC 2 O 4 + , and Am(C 2 O 4 ) 2 - ] to high ionic strengths, we have obtained the solubility constants (log 10 K 0 ) for the following reactions at 25 °C, Pr 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Pr 3+ + 3C 2 O 4 2- + 10H 2 O(l). Nd 2 (C 2 O 4 ) 3 •10H 2 O ⇌ 2Nd 3+ + 3C 2 O 4 2- + 10H 2 O(l). to be -30.82 ± 0.30 (2σ), and - 31.14 ± 0.35 (2σ), respectively. These values for can be directly applied to Pu 2 (C 2 O 4 ) 3 •10H 2 O, Am 2 (C 2 O 4 ) 3 •10H 2 O and Cm 2 (C 2 O 4 ) 3 •10H 2 O. The model established for actinide oxalates by this study provides the needed knowledge with regard to solubilities of actinide/REE oxalates at various ionic strengths, and is expected to find applications in many fields, including the geological disposal of nuclear waste and the mobility of REE under the surface conditions, as Pr 2 (C 2 O 4 ) 3 •10H 2 O and Nd 2 (C 2 O 4 ) 3 •10H 2 O can be regarded as the pure Pr and Nd end-members of deveroite, a recently discovered natural REE oxalate with the following stoichiometry, (Ce 1.01 Nd 0.33 La 0.32 Pr 0.11 Y 0.11 Sm 0.01 Pb 0.04 U 0.03 Th 0.01 Ca 0.04 ) 2.01 (C 2 O 4 ) 2.99 •9.99H 2 O. Regarding its importance in the geological disposal of nuclear waste, Am 2 (C 2 O 4 ) 3 •10H 2 O/Pu 2 (C 2 O 4 ) 3 •10H 2 O/Cm 2 (C 2 O 4 ) 3 •10H 2 O can be the source-term phase for actinides, as demonstrated by the instance in the disposal in clay/shale formations. This is exemplified by the stability of Am 2 (C 2 O 4 ) 3 •10H 2 O in comparison with Am(OH) 3 (am), Am(OH) 3 (s) and AmCO 3 (OH)(s) under the relevant geological repository conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

International collaboration, the route to fuel cycle research and development

In hindsight, involvement with European Framework projects such as GoFastR (Gas-cooled Fast Reactors) and ACSEPT (Actinide Recycling by Separation and Transmutation) was a crucial and, at the time, an innovative step in maintaining the UK skills base during a period of major changes in the UK nuclear industry. It has undoubtedly delivered the objectives intended in terms of maintenance of the key skills, developing and training new staff, regenerating facilities and building strong links with the European nuclear research community. Over the last 2-3 years NNL's participation in European projects has moved forward such that NNL (National Nuclear Laboratory) is an integral partner of several major projects, fully engaged with delivering the core objectives of the projects and intent on forging deep collaborations with key organisations across Europe. With the renewed interest in nuclear energy and future fuel cycle options in the UK, NNL is now well positioned to contribute at an even deeper level in European level programmes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

SiC Schottky Diode Detectors for Measurement of Actinide Concentrations from Alpha Activities in Molten Salt Electrolyte

In this project, we have designed a 4H-SiC Schottky diode detector device in order to monitor actinide concentrations in extreme environments, such as present in pyroprocessing of spent fuel. For the first time, we have demonstrated high temperature operation of such a device up to 500 °C in successfully detecting alpha particles. We have used Am-241 as an alpha source for our laboratory experiments. Along with the experiments, we have developed a multiscale model to study the phenomena controlling the device behavior and to be able to predict the device performance. Our multiscale model consists of ab initio modeling to understand defect energetics and their effect on electronic structure and carrier mobility in the material. Further, we have developed the basis for a damage evolution model incorporating the outputs from ab initio model in order to predict respective defect concentrations in the device material. Finally, a fully equipped TCAD-based device model has been developed to study the phenomena controlling the device behavior. Using this model, we have proven our concept that the detector is capable of performing alpha detection in a salt bath with the mixtures of actinides present in a pyroprocessing environment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Spent Nuclear Fuel Canister Capacity: Have the Limits Been Reached?

The commercial US nuclear utilities have been transferring spent nuclear fuel (SNF) from spent fuel pools (SFPs) to dry storage systems since 1986. SFPs were initially constructed for temporary storage, but without SNF reprocessing options in the United States and with no final disposal pathway yet available,SFPs started to reach capacity. Hence, to allow continued operation of the nation’s commercial nuclearreactor fleet, utilities started transferring SNF from SFPs (wet storage) to dry storage. Dry storage systems have evolved from storage-only systems first loaded in 1986 to the high-capacity dual-purpose storage and transport canisters used today. A question has arisen as to whether maximum canister capacity has been reached at the current capacity of ~50 kW heat load for storage, with 37-assembly pressurized water reactor and 89-assembly boiling water reactor canisters. This paper presents a holistic overview of commercial US SNF management practices over the years, the current SFP status in terms of filled (occupied) and empty (available) locations, the expected future reactor discharges, and the decay heat load distribution of SFP inventories as of June 2013. Finally, the paper makes a case that further increase in dry storage system capacity both in terms of number of assemblies and decay heat (1) is not needed, and (2) may not be practical from the cost-benefit perspective. b

Banerjee, Kaushik↗
DOE OSTI2017

Effect of Concentration, Temperature, and Interelectrode Gap on the Voltage Drop in an Electrochemical System of GdCl{sub 3}-LiCl-KCl

One of the most common conversations for those in the nuclear field and the public when discussing nuclear power is the issue of nuclear waste. When used nuclear fuel (UNF) is removed from a reactor, it contains fission products, some of which are long lived isotopes that will continue to irradiate for hundreds of years. One of several solutions to this issue is reprocessing, a process where minor actinides and other fission products are separated from still fissile and fertile materials. In the United States, there is no reprocessing of UNF at an industrial scale; however, one of the methods that universities and national labs are currently researching is known as pyro-processing. Pyro-processing utilizes electrochemistry in a molten salt environment to treat UNF and remove uranium and plutonium. The heart of the pyro-processing system is an electrorefining process, where most of the U in the UNF is selectively collected on a solid cathode in molten lithium-potassium chloride (LiCl-KCl) at high temperatures (723 K - 773 K). To model the behavior of materials in the electro-refiner, small bench-scale electrochemical experiments are being performed at many research universities and national labs to better understand material behavior. Within the electrochemical cell, there is a small resistance to current between the cathode and anode. This results in a voltage drop, which can affect the results of the experiments. This voltage drop varies based upon concentration of the salt, the temperature, the distance between the anode and cathode, and the scan rate. For this study, cyclic voltammetry will be performed on GdCl{sub 3}-LiCl-KCl at various temperatures (723 K - 798 K), various sample concentrations (1, 5, and 10 wt%), various scan rates (50, 100, 200, and 300 mV/s), and two different electrode spacing configurations. The resulting cyclic voltammograms will be compared for both electrode configurations to see the impact of interelectrode gap on the voltage drop.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Generation and Study of Am(IV) by Temperature-Controlled Electron Pulse Radiolysis

Used nuclear fuel (UNF) separation techniques that strive to separate radiotoxic americium (Am) from trivalent lanthanide fission products through oxidation state control have increased research efforts surrounding Am(V) and Am(VI). However, equivalent knowledge of the tetravalent state, Am(IV), has remained elusive, particularly in conditions more representative of UNF reprocessing, i.e., in concentrated nitric acid (HNO3). With this in mind, we have used electron pulse radiolysis to study the radiation-induced redox reaction of Am(III) with the oxidizing nitrate radical (NO3?) in 6 M HNO3: Am(III) + NO3? ? Am(IV) + NO3? . These experiments enabled us to observe the growth and decay of Am(IV) in a concentrated acidic solution for the first time. The transient Am(IV) species was found to have a lifetime of ~16 µs?sufficiently long-lived to play a critical mechanistic role in UNF reprocessing systems. Additionally, we performed the first-ever temperature-dependent kinetics study of an actinide element, elucidating unprecedented Arrhenius and Eyring activation parameters for the reaction of Am(III) with NO3?. This new knowledge provides much-needed molecular-level insights into the radiation-induced behavior of Am.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
NASA NTRS1977

Design and calculated performance and cost of the ECAS Phase II open cycle MHD power generation system

A 2000 MWe MHD/steam plant for central station applications has been designed and costed as part of the Energy Conversion Alternatives Study (ECAS). This plant is fueled by Illinois No. 6 coal, rejects heat through mechanical draft wet cooling towers, and includes coal processing equipment, seed reprocessing, electrical inversion of the MHD generator output and emission controls to current EPA standards. It yields an estimated overall efficiency of 0.483 (7066 Btu/kWe-hr), a capital cost of $718 per kWe (1975 dollars), and a cost of electricity at 65% capacity factor of 32 mills per kWe-hr. If the assumed life and reliability could be achieved with these performance parameters, the MHD system should prove attractive.

Harris, L. P.↗
DOE OSTI2021

Fuel Fabrication and Single Stage Aqueous Process Modeling

The Material Protection, Accounting, and Control Technologies program utilizes modeling and simulation to assess Material Control and Accountability (MC&A) concerns for a variety of nuclear facilities. Single analyst tools allow for rapid design and evaluation of advanced approaches for new and existing nuclear facilities. A low enriched uranium (LEU) fuel conversion and fabrication facility simulator is developed to assist with MC&A for existing facilities. Measurements are added to the model (consistent with current best practices). Material balance calculations and statistical tests are also added to the model. In addition, scoping work is performed for developing a single stage aqueous reprocessing model. Preliminary results are presented and discussed, and next steps outlined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Complete Initial Evaluation of Novel Complexants for Tc Holdback for Simplified Single Cycle Separations.

Management of technetium at the back end of nuclear fuel cycle is complicated by its unique physico-chemical character, but must be addressed to due to the environmental, storage and reprocessing challenges of this fission product. In a single-cycle scenario, the recovery and purification of uranium is particularly challenging due to the tendency of a pertechnetate anion to follow uranyl cation. This report summarizes initial studies directed towards finding the solution for technetium management. Two options for controlling technetium’s chemistry in solvent extraction scenarios were studied. Molecular recognition of pertechnetate by aqueous complexants based on guanidinium moieties was compared to redox manipulation of technetium using dihydrazide reagents. Solvent extraction, potentiometric and nuclear magnetic resonance spectroscopy studies were performed. The influence of guanidinium reagents on the liquid-liquid partitioning of technetium is less pronounced, relative to dihydrazides. Strong impact exerted by dihydrazides originates from the technetium-catalyzed decomposition of such compounds which reduces technetium to a tetravalent, non-extractable state. Although this route of technetium management is very effective the destruction of dihydrazides is undesirable. Guanidinium complexants show a more tempered influence on technetium. The effect on the liquid-liquid partitioning of pertechnetate is evident and likely guided by anion recognition due to ion-pair and hydrogen bond formation. Guanidiniums are stable in presence of technetium as evidenced by nuclear magnetic resonance studies. The steady state chemistry of guanidiniums identify this class of aqueous complexants as solid candidates for structure-function pertechnetate recognition studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

An assessment of the effect on Olkiluoto repository capacity achievable with advanced fuel cycles

Previously a few scenarios have been simulated for transition from thermal to fast reactor fleet in Finland in order to determine how much the transuranic inventory could be reduced with the partitioning and transmutation (P-T) technologies. Those calculations, performed with COSI6 code developed by CEA, are extended in the present study, in which the effect of P-T on the capacity of the planned final disposal repository at Olkiluoto (Finland) is evaluated by taking into account the created fission products and transuranic residuals from the reprocessing operations. The decay heat is assumed to be the most restrictive factor in defining the waste disposal packing density. The repository capacity evaluation of this study is based on the comparison of the decay heats produced by the deposited waste in various scenarios. The reference scenario of this article involves only Light Water Reactors (LWR) in an open fuel cycle. The capacity requirement of the geological repository is estimated in a few closed fuel cycle scenarios, all including actinide transmutation with Fast Reactors (FR). The comparison between the P-T scenarios and reference is based on the decay heat production of the deposited waste. The COSI6 code is used for simulations to provide the repository decay heat curves. Applying the closed fuel cycle would change the disposal concept and schedule, because of which it is not quite straightforward to assess the impact of P-T on the capacity. However, it can be concluded that recycling the transuranic nuclides probably decreases the required volume for the disposal, but thermal dimensioning analysis is needed for more specific conclusions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Initial calculations for source term of Molten Salt Reactors

This paper provides an overview of the current MSR design space and lists unique features of the various designs under consideration. Some general considerations for source terms calculation for Molten Salt Reactors (MSRs) are explained. Applicability and limitations of terminology currently defined for legacy light water reactor (LWR) systems are discussed in the view of MSRs and the need for updated terminology is discussed. Calculations carried out for the Molten Salt Reactor Experiment (MSRE) are discussed with a qualitative comparison to the designs presented. The nature of the fission products (FPs) and actinides for Low enriched uranium, thorium and fast U/Pu fuel cycles employed in representative molten salt reactor systems are discussed. Computational results are obtained from a code (Serpent 2) with online reprocessing. Divergence in source terms when fission product bubbling is demonstrated. The source release for each molten salt reactor during postulated accidents is also presented.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

DISTRIBUTION OF LANTHANIDE AND ACTINIDE ELEMENTS BETWEEN BIS-(2-ETHYLHEXYL)PHOSPHORIC ACID AND BUFFERED LACTATE SOLUTIONS CONTAINING SELECTED COMPLEXANTS

With the renewed interest in the closure of the nuclear fuel cycle, the TALSPEAK process is being considered for the separation of Am and Cm from the lanthanide fission products in a next generation reprocessing plant. However, an efficient separation requires tight control of the pH which likely will be difficult to achieve on a large scale. To address this issue, we measured the distribution of lanthanide and actinide elements between aqueous and organic phases in the presence of complexants which were potentially less sensitive to pH control than the diethylenetriaminepentaacetic (DTPA) used in the process. To perform the extractions, a rapid and accurate method was developed for measuring distribution coefficients based on the preparation of lanthanide tracers in the Savannah River National Laboratory neutron activation analysis facility. The complexants tested included aceto-, benzo-, and salicylhydroxamic acids, N,N,N',N'-tetrakis(2-pyridylmethyl)ethylenediamine (TPEN), and ammonium thiocyanate (NH{sub 4}SCN). The hydroxamic acids were the least effective of the complexants tested. The separation factors for TPEN and NH{sub 4}SCN were higher, especially for the heaviest lanthanides in the series; however, no conditions were identified which resulted in separations factors which consistently approached those measured for the use of DTPA.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Discrete-event Simulation Process Model for the Pyrochemical Processing of Plutonium at Los Alamos National Laboratory

The pyrochemical metal production operations that occur in the Plutonium Facility at Los Alamos National Laboratory perform plutonium purification with the aim to provide plutonium metal for a variety of defense- and non-defense missions within the National Nuclear Security Administration. The demands and constraints associated with the pyrochemical processing of plutonium are complex, making decision analyses challenging for program managers who require plutonium production for their mission applications. The construction of a discrete-event simulation process model is proposed to measure and report the process capacity, material throughput, equipment requirements, and dose accumulation for operators of the pyrochemical metal production operations. The process model, constructed in the ExtendSim™ software, will represent the cause-and-effect relationships between the pyrochemical processing environment and the process constraints, including criticality limitations, material control and accountability measures, chemical analysis requirements, and equipment availability. An accurate representation of the pyrochemical metal production process capacity through simulation modeling will be helpful to program managers in their efforts to forecast plutonium availability for mission applications. Furthermore, the proposed process model will be vital for future analyses that will measure the interactions between the pyrochemical metal production operations and the aqueous reprocessing operations and their ability to minimize transuranic waste disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Sensitivity and Uncertainty Quantification of Transition Scenario Simulations

This report documents the first collective attempt at developing and applying capabilities to quantify uncertainties, assess parametric sensitivities, and optimize multiple parameters and metrics in fuel cycle simulations generated by the SA&I Campaign. To do this, external codes that were designed to perform sensitivity analysis and uncertainty quantification (SA&UQ) needed to be coupled to the SA&I Campaign’s nuclear fuel cycle simulators (NFCS). In FY20, two approaches were pursued: 1) coupling Cyclus to an ORNL-internal code called MOT (Metaheuristic Optimization Tool) and 2) coupling DYMOND to the opensource SA&UQ tool kit Dakota. The primary objective of having these NFCS/SA&UQ coupled capabilities is to better inform DOE-NE and other stakeholders on the results generated from the NFCS. For a given set of fuel cycle strategies, policies, and technology assumptions that make up a fuel cycle scenario, these NFCS have traditionally been used by the SA&I Campaign to provide quantitative answers in terms of year-by-year mass flows, infrastructure requirements, costs, etc. With these newly developed coupled capabilities, the SA&I Campaign can now efficiently simulate hundreds or thousands of these scenarios, sample large ranges of parameters and assumptions, and use the unique features of the SA&UQ tools to process the data. This enables providing answers with known and propagated uncertainties, determining the sensitivity of important metrics to different parameters and assumptions, quantifying how much fuel cycle and technology parameters impact each other, and producing optimized fuel cycle strategies for single and multiple variables. To demonstrate these new capabilities, the Cyclus/MOT was used to model several scenarios ranging from simple fleet retirements to transitions to advanced reactors. Specifically, for a transition scenario from LWRs to SFRs and advanced LWRs, uncertainty quantification, sensitivity analysis, and optimization studies were applied to cases involving single and multiple parameter (input) and single and multiple metric (output) variations. In addition, a similar transition scenario was modeled to demonstrate how to optimize the reprocessing capacity parameter to minimize two performance metrics while taking into account uncertainties from two other parameters. Lastly, a depletion module based on SCALE/ORIGEN was added in Cyclus to simulate the third scenario that was designed to quantify the impact of the modeling assumption that all LWR used nuclear fuel have the same burnup. The newly developed DYMOND/Dakota capability was also applied to a transition scenario from the existing fleet to small modular reactors and fast reactors. This particular scenario involves not only explicit isotopic depletion via ORIGEN-2, but also includes multirecycling and utilizing the criticality search feature to determine the fresh fuel composition of recycled fuel, a feature unique to the DYMOND NFCS. A large database of simulations were run with 4 main parameters that were sampled: start date of reprocessing, reprocessing capacity, energy demand growth rate, and advanced reactor share of the fleet. The 4 main metrics were uranium consumption, enrichment requirements, waste generation, and levelized cost of electricity using data from the Cost Basis Report. The demonstrated SA&UQ results include those that inform on how to choose parameters to avoid “failed” scenarios, Sobol’ indices that inform on the importance of various parameters individually and synergistically, and Analysis of Variance (ANOVA) studies that decompose parameter ranges into groups and informs on whether variations are statistically significant.

Feng, B.↗
DOE OSTI2013

Nuclear power fleets and uranium resources recovered from phosphates

Current light water reactors (LWR) burn fissile uranium, whereas some future reactors, as Sodium fast reactors (SFR) will be capable of recycling their own plutonium and already-extracted depleted uranium. This makes them a feasible solution for the sustainable development of nuclear energy. Nonetheless, a sufficient quantity of plutonium is needed to start up an SFR, with the plutonium already being produced in light water reactors. The availability of natural uranium therefore has a direct impact on the capacity of the reactors (both LWR and SFR) that we can build. It is therefore important to have an accurate estimate of the available uranium resources in order to plan for the world's future nuclear reactor fleet. This paper discusses the correspondence between the resources (uranium and plutonium) and the nuclear power demand. Sodium fast reactors will be built in line with the availability of plutonium, including fast breeders when necessary. Different assumptions on the global uranium resources are taken into consideration. The largely quoted estimate of 22 Mt of uranium recovered for phosphate rocks can be seriously downscaled. Based on our current knowledge of phosphate resources, 4 Mt of recoverable uranium already seems to be an upper bound value. The impact of the downscaled estimate on the deployment of a nuclear fleet is assessed accordingly. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Tensioned metastable fluid detector sensing technology for multifarious-multiscale applications in the nuclear fuel cycle

The nuclear fuel cycle presents formidable 10 30+ multiscale challenges for high-efficiency spectroscopic sensing of alpha-neutron-fission radiation signatures amidst extreme gamma-beta radiation backgrounds at various stages of the cycle; it involves diverse arenas ranging from continuous/pulsed power production, security/safeguards, spent fuel, environmental remediation, and personnel health-dosimetry. This paper discusses the novel tensioned metastable fluid detector (TMFD) sensor technology. It discusses the results of studies demonstrating extreme (10 4 Sv/h) gamma-beta field blind high intrinsic efficiencies (80–100%) for neutron-alpha-fission radiation spectrometric detection. Example studies discussed in this paper include Neutron detection with spectroscopy; H*10 neutron radiation dosimetry upwards from ultra-low10 −2 µSv/h levels; Air and/or liquid-borne trace (10 –3 Bq/mL) quantity, 1 keV energy resolved alpha/fission spectroscopy for Rn/Po/U/Pu/Am environmental isotope assays at 106:1 (β:α) ratios; Directional-position enabled extended standoff tracking of neutron emitting (shielded/unshielded) special nuclear materials while stationary or moving at highway speeds; and, U.S. Department of Energy nuclear complex wide nuclear waste reprocessing, criticality safety, health physics, and environmental applications.

Alpha-fission-neutron spectrometry↗
DOE OSTI2020

Effect of ionizing radiation on PuO2 and surrogate oxide materials under storage conditions

The stockpile of plutonium separated from used nuclear fuel amounts to many 100s of tonnes. The majority of this plutonium is stored as PuO2 powder packaged in multi-can containers. The conditions during the packaging of reprocessed PuO2 for long term storage are closely controlled to limit water uptake as radiolytic decomposition of any adsorbed water will lead to the formation of both a potentially flammable atmosphere containing molecular hydrogen as well as reactive oxygen species which may be incorporated into the oxide phase. Consequently, the safety case for the long-term storage of PuO2 requires a complete understanding of the fundamental physical, chemical and materials degradation processes occurring at the water-oxide interface inside the storage canisters. This study investigates the effect of ionizing radiation on PuO2 and various surrogate oxides under atmospheres potentially encountered in storage canisters.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗