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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Impact of f-element complexation on the radiolytic robustness of separations ligands

Impact of f-element complexation on the radiolytic robustness of separations ligands Gregory P. Horne, Makayla R. Baxter, Corey D. Pilgrim, Travis S. Grimes, Center for Radiation Chemistry Research, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID, 83415, USA Cristian Celis Barros, Department of Chemistry and Biochemistry, Florida State University, Tallahassee, FL 32306, USA. E-mail: ccelisbarros@fsu.edu Andrew R. Cook, Department of Chemistry, Brookhaven National Laboratory, Upton, New York, 11973, USA Stephen P. Mezyk, Department of Chemistry and Biochemistry, California State University Long Beach, 1250 Bellflower Boulevard, Long Beach, California, 90840-9507, USA The effects of ionizing radiation are ubiquitous throughout all aspects of a nuclear fuel cycle. However, the complexity and intensity of these effects are greatest during the management of used nuclear fuel, owing to the presence of a wide spectrum of radionuclides from neutron capture and fission processes. With regards to used nuclear fuel (UNF) reprocessing, radiation-induced processes typically promote the destruction of active compounds (e.g., complexants and additives) with the concomitant formation of potentially detrimental degradation products and corresponding changes in physical and chemical properties, which ultimately impact the effectiveness and longevity of a given reprocessing system. Concerning UNF complexants, radiation chemistry studies have historically focused on their radiation robustness in the absence of the metal ions they were designed to selectively complex. This knowledge gap is worrisome as previous studies on aqueous phase complexants have demonstrated significant changes in radiolytic behavior upon metal ion complexation.1-4 More recently, the rate of reaction of the n-dodecane radical cation—believed to be the major organic phase radiation-induced transient species responsible for complexant radiolysis in n-dodecane based solvent systems—with hexa-n-octylnitrilo-triacetamide (HONTA) was shown to increase by an order-of-magnitude upon complexation of europium or americium.5 These findings have significant implications on the projected longevity of complexants in UNF reprocessing solvent systems. Consequently, a thorough understanding of metal ion complexation effects on the radiolytic integrity of UNF complexants is essential to evaluate their potential for process application. Presented here are two recent studies from the Idaho National Laboratory Center for Radiation Chemistry Research group that demonstrate the various impacts of f-element complexation (uranium, americium, and lanthanides) on the radiolytic integrity (gamma and electron pulse) of tributyl phosphate (TBP), N,N-di-(2-ethylhexyl)butyramide (DEHBA), N,N-di-2-ethylhexylisobutryamide (DEHiBA), and 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) under UNF reprocessing conditions. References 1) Bhattacharyya and Kundu, Int. J. Radiat. Phys. Chem., 1971, 3, 1. 2) Kundu and Matuura, Int. J. Radiat. Phys. Chem., 1975, 7, 565. 3) Ilan and Czapski, Biochimica et Biophysica Acta, 1977, 498, 386. 4) Buettner, Doherty, and Patterson, Fed. Euro. Biochem. Soc., 1983, 158 (1), 143. 5) Toigawa, Peterman, Meeker, Grimes, Zalupski, Mezyk, Cook, Yamashita, Kumagai, Matsumura, Horne, PCCP, 2021, 23, 1343.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Modeling the Fundamental Radiation Chemistry of the Organic Diluent, and the Effect of Metal Ion Complexation on the Radiochemical Behavior of Active Compounds

All used nuclear fuel (UNF) reprocessing technologies must operate efficiently in the presence of an intense, multi-component (predominantly alpha, beta, and gamma) radiation field. Consequently, radiation-induced degradation of reprocessing systems is of concern, as it negatively impacts process performance over time due to the destruction of both active compounds (ligands, phase modifiers, holdback agents, etc.) and the formation of degradation products. Reprocessing solvent system radiolysis has been linked to changes in separation efficiency and physical properties of solvent mixtures, solvent-recycle longevity, crud formation, and other unexpected outcomes that impact the efficient recovery of valuable materials (e.g., the actinides) and the volume of hazardous radioactive waste for final disposal, i.e., in a geological repository. Consequently, a fundamental understanding of radiolytic processes and their effects on reprocessing solvent system performance is critical for: (i) the cost-effective development and innovation of separation technologies; (ii) the design and implementation of predictive radiation chemical models for process monitoring and lifetimes; and (iii) potentially the ability to exploit radiolytic phenomenon to our benefit, e.g., strategic radiolysis of active molecules to liberate specific degradation products that aid subsequent process stages. Despite extensive investigation into the radiolytic behavior of active solvent system compounds, little attention has been given to understanding (i) the radiation chemical behavior and modification of the organic diluent and (ii) the effect of metal ion complexation on the radiochemical behavior of active compounds.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Organic Iodine Adsorption by AgZ under Prototypical Vessel Off-Gas Conditions

U.S. regulations will require the removal of 129 I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. Multiple off-gas streams within a UNF reprocessing plant combine prior to release, and each of these streams contains some amount of iodine. For an aqueous UNF reprocessing plant, these streams include the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas and the shear off-gas. To achieve regulatory compliance, treatment of multiple off-gas streams within the plant must be performed. Preliminary studies have been completed on the adsorption of I 2 onto silver mordenite (AgZ) from prototypical VOG streams. The study reported that AgZ did adsorb I 2 from a prototypical VOG stream, but process upsets resulted in an uneven feed stream concentration. The experiments described in this document both improve the characterization of I 2 adsorption by AgZ from dilute gas streams and further extend it to include characterization of the adsorption of organic iodides (in the form of CH 3 I) onto AgZ under prototypical VOG conditions. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the effect of sorbent aging on iodine removal in VOG conditions could be inferred.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Distribution Development for Residual Inventory at the New York West Valley Site - 20400

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351 ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into Agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States. NFS, a private company, built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations ended in 1972 and never reopened, leaving behind radioactive and chemical wastes. Operations led to contamination in a number of facilities and locations. Some of that contamination has migrated from waste disposal zones to other layers, formations, and features on and off the WNYNSC. Phase I decommissioning activities are ongoing and involve the removal of a number of areas and structures that have been associated with contamination. The purpose of this work is to outline the approach for characterizing contamination not associated with disposed wastes, contaminated structures, or specific releases. In this work, the term, residual radiological activity, is used to describe environmental contamination that exists subsequent to the completion of Phase I decommissioning activities, that is not associated with disposed wastes, contaminated structures, or specific releases. Contamination from the Site was quantified relative to data that characterize the concentrations of radionuclides that exist in background. Background concentrations are those present in the area but having no influence from Site related activities. The existence of residual radiological activity that is elevated relative to background has the potential to contribute to future risks to human health and the environment. As a consequence, the residual inventory information is used to inform the West Valley Probabilistic Performance Assessment (PPA) model to characterize potential future risks to human health and the environment. The centralized West Valley Data Management System (DMS) was the source of information for the data assembled in this analysis. The DMS is a fairly large compilation consisting of thousands of records from investigation studies, with sample dates ranging from 1990 to present. Samples from monitoring wells, boreholes, geoprobe studies, surface water, surface soils, storm water outfalls, ventilation stack filters, plant and animal tissues, and more are included in the DMS. Results are typically reported in units of activity per unit volume. For the purpose of the analyses presented here, all results were converted into consistent units of pCi per unit volume. Since 1990, data have been collected from various locations across the WNYNSC at different times with varying frequency over the course of several decades. As a consequence, a number of potential issues can arise with respect to the assembly of a dataset that is deemed adequate for the characterization of residual radiological activity. These issues were assessed and resolved to the extent possible through careful consideration of the properties of the distributions. The intent was to use data which characterize the current state of the Site. Radionuclides can be designated to one of several groups depending on their origin. In this work the groups considered were 1) Naturally Occurring Radioactive Material (NORM), 2) fallout, and 3) Other (including power plant, medical research, etc). This grouping is a useful construct with respect to the interpretation of fixed laboratory results. For example, NORM radionuclides that exist within a decay chain should have approximately equivalent distributions of concentrations if they are representative of background conditions. Insights such as these can be used as a check to identify sample results that need to be further investigated or omitted due to issues associated with reported results from fixed laboratory analyses. This type of analysis provided a foundation for the assessment of the adequacy of sample results for use in subsequent components of an assessment. The general process for the assessment of residual radiological contamination at the Site consists of a sequence of several steps. First, for each analyte, several statistical tests were performed to assess the weight of evidence against the null hypothesis that the mean of the distribution of concentrations was equal to zero. If the mean of the distribution of concentrations for a given radionuclide was not found to be greater than zero, then it was removed from consideration as a component of the residual radiological contamination. If there was significant evidence to reject the hypothesis of the mean being equal to zero, the second step was to compare the distribution of the data from the Site to that of the corresponding background. A suite of tests was used to compare the distributions of the site and background data. The results of these tests were collectively used to determine if site data are elevated relative to background. The third step was to develop distributions using a Bayesian framework to characterize the distribution of mean of the increment present above background for each of the radionuclides. The Bayesian model implemented allowed for the comparison of site-specific records to background concentrations to better approximate contamination attributed to the Site. A final screening step was employed for radionuclides that exceed background. This screening step compared 95% upper confidence limits (UCLs) from the increment distribution developed in the previous step to the risk screening levels. This approach yields a list of analytes that were determined to be elevated relative to background.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Quantifying the impact of climate change on erosion - 20397

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha (3,338-ac) site located approximately 48 km (30 mi) south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into Agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States. NFS, a private company, built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (metric tons, or 705 short tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations ended in 1972 and never reopened, leaving behind radioactive and chemical wastes. Erosion can play an important role in the fate and transport of waste at sites where disposal of long-lived waste is anticipated. The statistical characterization of key processes related to erosion is essential to the understanding of site stability through time. One of the key processes governing erosion is extreme precipitation and it is critical that trends in the distribution of extreme precipitation events through time be represented. The evidence of climate change is increasingly well documented and projected impacts on extreme precipitation events should be incorporated in performance assessment studies when relevant. Not evaluating future climate states in a performance assessment is contradictory to good modeling practice. Specifically, excluding climate change limits development of modeling information that could aid in effective decision making. The current climate literature provides both observational evidence and climate model projections of climate trends and/or climate change in the late 20. and early 21. centuries for North America and the northeast United States. In this work, this information was used to assess the impacts of potential changes in climate on erosion processes. The goal was to understand how projections of future climate relate to the performance of the WNYNSC through time. In the first stage of this work multi-temporal historical aerial images were analyzed in conjunction with orthophotography and Lidar data to develop probability distributions for variables representing important erosion processes. The information from these analyses was then used in conjunction with simulated data from the West Valley Erosion Working Group (EWG). Analysis of EWG simulations provides estimates for the change in erosion rates through time that is driven by changes in climate. The time-varying rates of erosion change were applied to the historical aerial imagery data in order to inform time-varying rates of erosion that are driven by changes in climate. Ultimately, this process resulted in the identification of locations for features like gully heads using both the Lidar dataset and projection of the estimated location from the historical aerial photo under consideration back to the Lidar dataset. The distance between the estimate of the location from the historical image and that of the Lidar dataset was the estimated distance the feature has moved. This distance was then divided by the number of years between the Lidar dataset and the year of the historical aerial image of interest to get a rate of movement through time. This was done for several dozen points on each historical aerial image. The analyses from the EWG were used to characterize the relative impact of climate change on the erosion rates. This relative impact was quantified by comparing the LEM based estimates of erosion that were derived using historical climate data with LEM based estimates of erosion that were derived using projections of future climate. The relative increase in the erosion rates was then applied to the historical estimates of erosion derived from the analysis of the historical aerial imagery. This approach was used since the LEM-based estimated of the historical erosion rates from the past century have a bias towards underestimating erosion. This underestimation is hypothesized to be a consequence of an inability of the LEMs to account for the impacts of land-use change (i.e. deforestation) which had been shown to significantly increase erosion in similar Northern hardwood forest ecosystems. In summary, gully head retreat rate and gully widening rate were characterized using statistical probability distributions from the historical aerial image analysis. Simulated data from the EWG was used to estimate the climatically-driven changes in erosion rates through time. The changes through time from the EWG were applied to the gully head retreat rate and gully widening rate characterized using the historical aerial image analysis. This information can be used to inform PPA models to project future risks from a given site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Structure of the Public Irradiation Dose During Operation of Experimental-Demonstration Power Complex Enterprises

The structure of the public radiation dose resulting from gas aerosol emissions of the experimental-demonstration energy complex incorporating the BREST-OD-300 reactor and fuel fabrication and reprocessing modules is determined. The yearly radiation dose to humans at the point of maximum ground concentration of radionuclides from design basis emissions of the energy complex is formed primarily by {sup 3}H,{sup 14}C, and fission products (0.73, 1.2 2, and 0.9 μSv, respectively) through the peroral pathway as a result of the fallout during the running year of operation of the enterprises. The largest contribution to the irradiation dose is due to the emissions from the spent fuel reprocessing module – 2.33 μSv/yr – and 50% is due to {sup 14}C. The radiation dose from the emissions of the BREST-OD-300 reactor is almost completely due to {sup 3}H and {sup 210}Po (0.73 and 0.17 μSv/yr, respectively). The emissions from the fuel fabrication module have the smallest effect on the public. The dose from fission products is produced by emissions from the reprocessing module approximately in equal amounts in terms of external and internal pathways (0.47 and 0.43 μSv).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Influence of elevated temperature and oxygen on the capture of radioactive iodine by silver functionalized silica aerogel

Reprocessing is considered a competent strategy for spent nuclear fuel management, yet radioiodine ( 129 I) is emitted in reprocessing off-gas as a hazardous byproduct. Silver functionalized silica aerogel (Ag 0 -aerogel), a promising iodine capture material, experiences a reduction in its capacity after prolonged exposure to off-gas components at elevated temperatures, a phenomenon termed as aging. To fully understand this process, we isolated the contribution of each aging factor, exposing Ag 0 -aerogel samples to N 2 and dry air gas streams, respectively, at 150 °C for different time periods. Aged samples were loaded with I 2 to examine the capacity change and comprehensively characterized to investigate the evolution of their properties. Results show that temperature alone did not alter Ag 0 -aerogel's capacity but triggered Ag 0 nanoparticles sintering and generated organic sulfur species. The presence of O 2 reduced the capacity by ~20 %, causing (i) formation of silver sulfide (Ag 2 S) crystals and (ii) oxidation of Ag-thiolate (Ag-S-r) to Ag sulfonate (Ag-SO 3 -r). Given that Ag 2 S readily adsorbs I 2 , the formation of Ag-SO 3 -r is the major inhibitor for iodine adsorption. This hypothesis was supported by density functional theory (DFT) simulations. These findings unraveled key mechanisms of Ag 0 -aerogel aging, which are useful in the development of materials that withstand realistic spent-nuclear-fuel-reprocessing off-gas conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Study on the recycling of waste PVC compounds from electrical wires

The good mechanical, thermal and electrical properties of poly (vinyl chloride) (PVC) make it a polymer used in many applications, among which is the coating of wires and cables. The processing of PVC for this application is made using the extrusion process. The PVC waste resulting from extrusion can be recycled or placed on landfills. The objective of this work is to develop recycling methods to decrease the amount of PVC stored in landfills. This work evaluated the influence of successive processing of PVC on the final properties of the obtained compound, which will be used in the process of coating wire and cable by extrusion. The mechanical properties of the electrical wires were assessed according to the Brazilian Association of Technical Standards (ABNT NBR 6251). Tests of tensile strength, elongation at break, before and after thermal aging in an oven with air circulation for a period of 168 hours at 100 °C were performed. The degradation of the PVC compound was carried out through Fourier transform infrared spectroscopy (FTIR) and mechanical properties evaluation. The results showed that the compound can be reprocessed five times keeping the minimum requirements of the ABNT NBR 6251 Standard. Reprocessing was stopped upon color changing (yellowing) of the compound. Yellowing is a characteristic degradation behavior by dehydrochlorination of PVC. FTIR analysis showed changes at 1600 cm{sup −1}, 1430 cm{sup −1}, 685 cm{sup −1} and 614 cm{sup −1} bands. The performance of the PVC compound for coating electrical wire and cable was fair, meeting the requirements of the ABNT NBR 6251 Standard even after five reprocessing cycles.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium–plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and {sup 238}U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with {sup 238}U alone should be reprocessed in the usual way.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Iodine Sorption on AgZ-PAN

The Department of Energy (DOE) Nuclear Fuel Cycle and Supply Chain Program Material Recovery and Waste Form Development (MRWFD) Campaign Off-gas Sigma Team has performed research and development on iodine control and iodine waste forms for the past several years. This research and development has included iodine adsorption tests using a laboratory-scale test system containing multiple-segmented fixed beds of iodine adsorbents. Iodine capture performance has been tested using (a) non-radioactive synthetic gas mixtures blended from air, nitrogen, NO, NO 2 , water, to simulate dissolver off-gas (DOG) and vessel off-gas (VOG) streams from used nuclear fuel aqueous reprocessing, (b) inorganic and organic iodine species, and (c) different sorbents. An initial test of silver-functionalized mordenite in polyacrylonitrile matrix (AgZ-PAN) was conducted to assess if this sorbent, initially designed for selective sorption of Xe that evolves into used fuel reprocessing off-gas streams, could also be a candidate for iodine sorption. This test was conducted using the same deep-bed test system and methodologies used in many prior iodine sorption tests using other sorbents. Diatomic iodine (I2) was used as the target iodide in a gas stream containing air, NO, NO 2 , and moisture designed to simulate an aqueous reprocessing dissolver off-gas (DOG) stream.

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

New Physical Insights about Tidal Disruption Events from a Comprehensive Observational Inventory at X-Ray Wavelengths

We perform a comprehensive study of the X-ray emission from 70 transient sources that have been classified as tidal disruption events (TDEs) in the literature. We explore the properties of these candidates, using nearly three decades of X-ray observations to quantify their properties and characteristics. We find that the emission from X-ray TDEs increase by two to three orders of magnitude, compared to pre-flare constraints. These emissions evolve significantly with time, and decay with power-law indices that are typically shallower than the canonical t {sup −5/3} decay law, implying that X-ray TDEs are viscously delayed. These events exhibit enhanced (relative to galactic) column densities and are quite soft in nature, with no strong correlation between the amount of detected soft and hard emission. At their peak, jetted events have an X-ray to optical ratio ≫1, whereas non-jetted events have a ratio ∼1, which suggests that these events undergo reprocessing at different rates. X-ray TDEs have long T {sub 90} values, consistent with what would be expected from a viscously driven accretion disk formed by the disruption of a main-sequence star by a black hole with a mass <10{sup 7} M {sub ⊙}. The isotropic luminosities of X-ray TDEs are bimodal, such that jetted and non-jetted events are separated by a “reprocessing valley” that we suggest is naturally populated by optical/UV TDEs that most likely produce X-rays, but this emission is “veiled” from observations due to reprocessing. Our results suggest that non-jetted X-ray TDEs likely originate from partial disruptions and/or disruptions of low-mass stars.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2021

Acid Dissolution Behavior of Ferritic FeCrAl Tubes Candidates for Nuclear Fuel Cladding

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel in light water reactors. One solution is to replace the zirconium cladding using ferritic iron-chromium-aluminum (FeCrAl) alloys, which offer extraordinary resistance to high-temperature reaction with air or steam due to the formation of a protective alumina layer on the external surface. It is important to characterize the behavior of FeCrAl not only during accident conditions but in the entire fuel cycle, which may include reprocessing of the used fuel after it is removed from the power reactors. The reprocessing may involve the dissolution of the fuel rods in mineral acids. Little or nothing is known on the dissolution of FeCrAl alloys in common mineral acids, therefore the objective of this research was to study the dissolution of typical cladding tubing having two compositions of FeCrAl (APMT and C26M) in three acids (H 2 SO 4 , HNO 3 , and HCl) as a function of the temperature using both standard ASTM immersion tests as well as electrochemical tests. The dissolution behavior of the FeCrAl alloys is compared to the dissolution capability of other traditional nuclear materials such as austenitic stainless steels (304SS and 316SS) and austenitic nickel alloys (Alloy 600 and Hastelloy C-276). Results show that both C26M and APMT have a higher dissolution capability in the studied mineral acids, which will be beneficial for reprocessing procedures.

36 MATERIALS SCIENCE↗
DOE OSTI2017

The Reactor Graphite Handling. Problems and Solutions - 17208

The problem of industrial uranium-graphite reactors and NPP units decommissioning is one of the urgent tasks that require efficient and cost-effective solutions. To date, only Russia has already shut down and subject to decommissioning more than 120 objects. World-wide, more than 260,000 ton of irradiated graphite have arisen from the commercial nuclear power operations, research reactors, and defense production reactors. Under the circumstances the problem to develop cost-effective methods of the irradiated graphite management has appeared. Alternative disposal options for the irradiated graphite were discussed for at least 25-30 years. To date, various methods of irradiated graphite reprocessing have been proposed. They are based on the application of different physical and chemical processes, most of which involve incineration of the entire mass of graphite. However, they all have disadvantages, and do not solve the whole complex of problems. The primary goal of the research presented here was to carry out a study aimed at the development of a more simple technology (as compared to the incineration of the entire mass of graphite) that would provide for the graphite purification and allow the possibility of its transfer to the less hazardous waste category followed by the disposal of it in a near-surface repository. To achieve it is proposed to consider two directions, each of them reducing the cost of graphite reprocessing and the waste volume disposed of in the repository. The first alternative is based on the principle of partitioning radionuclides of different nature from the graphite, thus allowing to transfer it into the LLW category and minimize the volume of secondary waste products. The second option to change the reprocessing technology is the idea to use reactor graphite as the starting material to produce stable final waste form compositions intended for the long-term storage of radioactive waste including transuranic elements. (authors)

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

EDF's used fuel management french policy - 15199

Regarding back-end fuel cycle, EDF's strategy, implemented from the very beginning of French nuclear production, consists in reprocessing of used fuel; decision to recycle plutonium in French PWR was taken from 1984. This reprocessing and recycling strategy, relying on existing industrial facilities (AREVA La Hague and Melox plants), has proven to be reliable and efficient. Treatment leads to a stabilization of used fuel total inventory in interim storage. Vitrification process ensures a safe packaging for high level and long-lived waste under a reduced volume. Recycling of fissile materials saves natural uranium requirements. Concentration of Pu in MOX used fuel leaves open the option to reuse it for starting next Generation IV systems beyond 2040. In 2008, EDF and AREVA, supported by French Government, agreed on commercial terms covering long term fuel reprocessing and recycling (up to 2040). Currently, EDF fuel management strategy relies on flow equilibrium in order to respect fuel cycle consistency and sustainability. The underlying principle is to strictly adjust separated plutonium production to recycling possibilities (i.e. licensed reactors and MOX fuel fabrication facilities). Given that fuel cycle policy, any trend in NPP core management that may impact the fuel cycle should be anticipated within times in order to avoid technical or administrative incompatibilities, and guarantee a high level of safety and robustness. In collaboration with industrial operators, EDF have to perform periodic 'Cycle Impact' analyses and report to the French Nuclear Safety Authority (ASN); next exercise is scheduled by the end of 2016. (authors)

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

Life Extension of Nuclear Plant Facilities through Monitoring and Detailed Analysis - 16175

The UK has accumulated a substantial legacy of radioactive waste from various civil nuclear and defence programmes. Presently, however most of the UK radioactive waste is generated from nine power stations which generate approximately one sixth of the of the country's electricity. Irradiated fuel is removed from these reactors and safe management of the radioactive solid and liquid effluents is required. The safe management of these effluents is essential. Sellafield, in West Cumbria UK, contains two reprocessing plants, one designed to process nuclear fuel from Magnox reactors (CO{sub 2} cooled with Uranium bar fuel) and the other to process nuclear fuel from Advanced Gas cooled Reactors and Pressurized Water Reactors. The reprocessing stage enables extraction of fissile material for recycling. This leaves a volume of dilute but high-level waste. The high-level radioactive liquid waste is concentrated to reduce its volume and then incorporated into borosilicate glass, using a process called vitrification. The glass is then poured into stainless steel containers and placed in long term storage. This required concentration of HLW liquor is achieved using a suite of aging evaporators. All evaporators are of similar design constructed from corrosion resistant stainless steel having dimensions height ∼11 m, internal diameter 3.05 m and capable of holding 13.7 m{sup 3} of liquor. The basic principle of operation is to heat the dilute liquor under partial vacuum, at constant volume, until the target concentration is achieved. Heating of the liquor is achieved using steam heated coils within the evaporator, and steam in the jacket which surrounds the lower half of the evaporator vessel. The evaporator is then brought to atmospheric pressure and the concentrated liquor is ejected whilst hot. The evaporator is then cooled with water cooled coils and a water filled jacket. New dilute liquor is then added for a new batch. The process is then repeated. As the evaporators reach the end of their original design life accurate prediction of the current condition of nuclear plant facilities is essential to safely justify continued plant operation by ensuring structural integrity. This poses a significant challenge as many aging UK nuclear facilities were not inherently built with inspection in mind. The potential life limiting components of the HA evaporators at Sellafield are the surfaces that contact the product liquor at elevated temperatures i.e. the steam heated surfaces of the coils within the evaporator vessel and the vessel base and wall surrounded by the steam jacket. Each heating component is critical to the operational life of the evaporator - the removal from service of a coil heating component significantly reduces the operational flexibility and remnant life of this plant. Over the last decade a variety of cutting-edge expertise and methodologies have been utilised to accurately measure and then predict the current condition of the plant and also to predict its remnant life. This paper provides an overview of the development of several techniques deployed, namely non-destructive testing/inspection, in-house software development and finite element modelling. These methods have been validated by lab-scale or full-scale experiments along with historical plant data. These techniques have been evolved and refined for over a decade, providing vital insights into the physical behaviour of this complex reprocessing system. This enables not only accurate assessment of the current condition of the plant equipment but also allows scientifically underpinned predictions to be made of how it will behave into the future. This is critical for Sellafield as this knowledge feeds into their safety case to the Office of Nuclear Regulation to justify continued operation. It is concluded that this integrated approach provides a robust analysis tool which has driven change and optimisation of plant procedures ensuring continued safe operation whilst extending the life of the facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Plans for the Removal of Inventory from HLW Storage Tanks and Impact on Vitrification Feeds - 18175

Over the next 4 years Sellafield will cease reprocessing nuclear fuel. This represents a significant change in the mission of Sellafield Ltd, shifting the site focus from reprocessing operations to preparations for decommissioning. The storage tanks which have stored the acidic high level waste produced during reprocessing will require washing out. These tanks have been used since the 1950's and over time the nature of this waste has changed. It poses unique challenges for both removal and vitrification. A large collaborative programme of work is underway at Sellafield to characterise this inventory and prepare a strategy for its removal and vitrification. (authors)

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

Going from Closed to Open and Then Back Again: How to Keep a Level Head

Nuclear Energy is a mature, reliable low carbon technology with a secure and abundant fuel source and is an essential contributor to the energy mix. The UK has a long history of using nuclear energy and has a long standing world class expertise in the recycling of used nuclear fuel with three generations of industrial scale reprocessing plant operating on the Sellafield site, as well as the Thorp Miniature Pilot Plant in the 1980's, and a vibrant, internationally recognised R and D capability at the National Nuclear Laboratory (NNL). However, the reprocessing plants at Sellafield are both scheduled to finish around 2020 with Thorp and Magnox plants moving to post operational clean out (POCO). This will move the UK from a closed fuel cycle to an open one. Within the UK a decision has been made to refresh the current nuclear energy plants to at least replacement of the existing nuclear fleet. This will mean the building of new nuclear power plant to ensure energy production of 16 GWe per annum. With the closure of the UK's reprocessing plants, these new nuclear power plants will operate on an open fuel cycle with long term storage of used fuel followed by disposal. However it is also recognised that this may not be enough and as such expansion scenarios ranging from 16 to 75 GWe nuclear energy production are being considered through the development of a UK Road map. Within these energy scenarios, a variety of options are being evaluated including electricity generation only, electricity generation plus heat, open versus closed fuel cycles, Generation III versus Generation IV systems and combinations of the above. What is clear is that the deciding factor on the type and mix of any energy programme will not be on technology choice alone. Instead a complex mix of Government policy, relative cost of nuclear power, market decisions and public opinion will influence the rate and direction of growth of any future energy programme. At the ANS Winter Meeting 2013 a number of these techniques were presented. What was clear from the fuel cycle scenario modelling was the requirement for the utilisation of a closed fuel cycle to enable the expansion beyond 16 GWe. It is therefore a necessity for the UK to keep options open for a potential return to a closed fuel cycle in the future. The requirement to maintain the skills and capabilities to do so will be an essential part of any future nuclear strategy. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗