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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Thermally and chemically responsive nanoporous materials for efficient capture of fission product gases.

The objective of this project was to develop robust, high-efficiency materials for capture of fission product gases such as He, Xe and Kr in scenarios relevant for both reactor fuels and reprocessing operations. The relevant environments are extremely harsh, encompassing temperatures up to 1500 °C, high levels of radiation, as well as potential exposures to highly-reactive chemicals such as nitric acid and organic solvents such as kerosene. The requirement for nanostructured capture materials is driven in part by the very short (few micron) diffusion distances for product gases in nuclear fuel. We achieved synthesis, characterization and detailed modeling of the materials. Although not all materials reviewed in this report will be feasible for the ultimate goal of integration in nuclear fuel, nevertheless each material studied has particular properties which will enable an optimized material to be efficiently developed and characterized.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Testing of an Integrated Iodine Scrubber and Polishing Bed System

In 2016 an initial engineering evaluation and design of the off-gas abatement systems required for a hypothetical 1000 t/yr used nuclear fuel (UNF) reprocessing facility treating 5 yr–cooled, 60 GWd/tIHM UNF was completed. In this study two types of silver-based iodine sorbents were considered. The first was a silver-exchanged mordenite (AgZ), and the second was a silver-functionalized silica aerogel. In the initial design of the dissolver off-gas (DOG) iodine capture beds, the sorbent capacity was assumed to be 40 mg I/g sorbent. The total iodine released from the fuel, used as input for the design of the DOG system, was 1.635 kg/day. Tramp halogens increased this to ~5.1 kg/day on an iodine equivalent basis. The daily consumption of iodine sorbent was ~166 kg (including the mass transfer zones that have lower iodine loading). The iodine capture system was designed to utilize five beds in parallel. In this design each bed was ~1 m in diameter and 1.5 m long. This design allowed the beds to remain on line for 20 days, but at that point all five beds would need to be replaced, resulting in considerable material handling. It was proposed that the bed life could be extended and the number of parallel beds reduced if an aqueous iodine scrubber preceded the solid sorbent iodine capture beds. The testing described in this report was intended to provide an evaluation of an integrated system in which bulk iodine would first be removed by an aqueous NaOH scrubber with a target iodine decontamination factor (DF) of 10 to 50. This would reduce the sorbent usage by 90 to 95% and significantly reduce the associated frequent remote handling of the iodine filter beds. The effluent from the scrubber would then be passed through an AgZ sorbent bed to remove residual iodine. The integration of these unit operations was intended to demonstrate an iodine DF of >1000, which would facilitate regulatory compliance with regards to total volatile radionuclide emission from a reprocessing facility. Initial baseline performance testing of the scrubber system includes the characterization of CO 2 removal by the aqueous NaOH scrubber as a function of temperature, caustic concentration, packing length, liquid-to-gas mass ratio, and liquid flow rate. It was found that when the column was well wetted, very high CO 2 DFs were achieved (>650 and potentially as high as 15,000), preventing the ability to elucidate the effects of each specific variable tested. The L/G ratio in this system was not representative of typical L/G ratios found in industrial-scale scrubber systems. By significantly lowering the liquid recirculation rate to the point that column wetting was marginal, some significant variations in the CO 2 DF or CO 2 effluent concentrations could be observed. The additional tests conducted under these more adverse conditions still showed CO 2 DFs >100. The only operational parameter with significant effects was determined to be column length, with increased column length resulting in higher CO 2 DFs. Following these CO 2 tests, a single iodine scrubber test was performed deliberately at a condition indicated by the CO 2 demonstration to result in poor column performance. This was intended to establish a lower performance bound for iodine removal by the aqueous scrubber. The result showed that even under the adverse operating conditions, an iodine DF of >100 could be achieved. Finally, the aqueous scrubber and a AgZ polishing bed were integrated. This simple caustic scrubber system achieved an iodine DF of >12,000 for just the scrubber portion. Using this iodine DF, the iodine loading on the polishing bed was reduced by 99.993%. The performance of that integrated system was compared to a system including only a AgZ sorbent bed. The system without the scrubber also showed iodine DFs >3000 until breakthrough was observed. The loading on the saturated portion of the bed averaged 87.2 mg I 2 /gm AgZ. Comparing the penetration of iodine into the sorbent beds for these two systems showed a marked difference; the penetration with the scrubber was only 0.65 cm at a maximum loading of 0.11 mg I 2 /gm AgZ, whereas for the case without the caustic scrubber, the bed was saturated for the first 7 cm and breakthrough of the 13.3 cm bed occurred.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Modeling and validating tritium transfer in a grassland ecosystem in response to {sup 3}H releases

In this paper a radioecological model (TOCATTA) for tritium transfer in a grassland ecosystem developed on an hourly time-step basis is proposed and compared with the first data set obtained in the vicinity of the AREVA-NC reprocessing plant of La Hague (France). The TOCATTA model aims at simulating dynamics of tritium transfer in agricultural soil and plant ecosystems exposed to time-varying HTO concentrations in air water vapour and possibly in irrigation and rain water. In the present study, gaseous releases of tritium from the AREVA NC nuclear reprocessing plant in normal operation can be intense and intermittent over a period of less than 24 hours. A first comparison of the model predictions with the field data has shown that TOCATTA should be improved in terms of kinetics of tritium transfer.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Management of HL deposits for the D and D of UP1 reprocessing plant - retrieval and treatment scenarios - 15295

Built in 1958 on the Marcoule CEA site, the UP1 facility, was the 1. French industrial (500 t U / year) reprocessing plant based on the PUREX process. The plant, which has stopped operation in 1997, constitutes with its auxiliary workshop (vitrification workshop, glass canister storage), one of the largest and most challenging decommissioning project in France. In about 40 years of operation, various type of fuels elements such as metallic uranium, U-Al fuel rods, Phenix breeder elements or PWR fuel have been reprocessed, thereby requiring several plant and process adaptations. Resulting from these complex process chemistry and operation history, significant fouling of the chemical plant circuits and components has occurred. The fission product (FP) concentration process is particularly concerned. Preliminary investigation performed in 2010 in the evaporator's feeding tanks revealed the presence of significant amounts of sludge and dose rates as high as 250 Gy/h. Therefore a thorough investigation campaign, including visual inspection, volume quantification, deposits sampling, and radiological characterization of the components of this workshop has been initiated in 2013. Chemical analysis performed so far show that the solids are mainly composed of zirconium phosphates, 'MoZr' compounds and nitrate salt precipitates. But the presence of insoluble fines arising from the dissolution workshop and organic residues resulting from the TBP decomposition has also been evidenced in some of the tanks. The numerous constraints associated with the management of such highly active sludge, with Cs-137 specific activity ranging from 1 to 10 GBq/g, and a complex physico-chemical composition, lead CEA to investigate the implementation of an in situ conditioning treatment able to produce a waste package complying with existing evacuation routes and/or on-site storage facilities. The full paper presents an overview of the UP1 D and D project and then focuses on the possible dismantling scenarios for the FP concentration workshop, including the evaluation of different strategies for the highly active deposits management. The investigations and deposits characterization program will be developed so as the sampling methodologies that has been specifically tuned. In the light of the best actual knowledge of the deposits properties, the recovery and conditioning treatment scenarios will be exposed, considering, as a case study, the development of a specific vitrification tool to be implemented In Situ and specifically designed to match the requirements and constraints of the D and D project: adequately dimensioned, compact, flexible with respect to the feed composition, for a limited duration of use and designed to be dismantled. (authors)

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

Review of on-line and near real-time spectroscopic monitoring of processes relevant to nuclear material management

Spectroscopic chemometric based on-line monitoring of used nuclear fuel (UNF) reprocessing solutions and characterization of legacy nuclear waste (LNW) stored at Hanford is discussed in this manuscript. Using on-line and near real-time monitoring with spectroscopy, as opposed to traditional off-line monitoring, can significantly reduce the cost, risk and improve the efficiency of characterizing UNF reprocessing streams and LNW since the data can be collected non-destructively and analyzed nearly instantly. It can accurately quantify and qualify nuclear material in near-real time which can help to know the immediate UNF reprocessing stream and LNW conditions and detect the diversion of materials. The ability to be applied in a real reprocessing plant and in an actual Hanford waste tank/transfer pipe has been demonstrated by applying this technique to accurately quantify analytes in real UNF streams and LNW samples. Furthermore, the future development of spectroscopy based on-line monitoring is also discussed in this manuscript.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Progress, Performance, and Prospects of Ultra-High Resolution Microcalorimeter Spectrometers

In 2005 the LANL/NIST team used a single high-resolution microcalorimeter detector to measure the gamma-ray spectrum of a plutonium sample. After more than a decade of research and development on this topic, both the technology and our general understanding of its capabilities have advanced greatly, such that a progress review is now timely. We examine the scenario of a large-scale reprocessing plant and conclude that current non-destructive analysis (NDA) methods are inadequate to safeguard such a facility to the desired levels, leading to undesirable dependence on massspectrometry (MS) destructive analysis (DA). The development of microcalorimeter detectors is intended to close the performance gap between NDA and DA methods to address the needs of nuclear facilities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Extended Elemental 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 environmental release, and each of these streams contains some amount of iodine. For an aqueous UNF reprocessing plant, these streams include the dissolver off-gas (DOG), the cell off-gas (COG), the vessel off-gas (VOG), the waste off-gas (WOG), and the shear off-gas (SOG). To achieve regulatory compliance, treatment of multiple off-gas streams within the plant must be performed. Of these remaining streams, the VOG is expected to be one of the major contributors to the balance of the residual iodine releases. The VOG will most likely contain iodine at parts-per-billion (ppb*) concentrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

IODINE ADSORPTION IN FIXED BED ADSORBER TESTS

I-129 emissions control is needed for used fuel aqueous reprocessing. I-129 capture has been studied for 40 years; Two sorbents tested in recent years – Silver Zeolite (AgZ) and silver-functionalized Aerogel (AGAero). Results show needed iodine adsorption efficiencies are possible, DFs up to10,000. Mass transfer zone (MTZ) depths ~2 in. for I2 and 8 in. for CH3I, but increase over time; deeper sorbent beds needed for efficient CH3I capture. Ag utilization between 10-70% for most test durations; lower for higher NOx. Iodine sorption capacities: ~10 g/100 AgZ and ~20 g/100 g AgAero; lower for high NOx. Nearly all adsorbed I was physisorbed iodine for I2 and CH3I.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Decontamination Process of Internal Part Pipes - 13442

The Marcoule Site, created in 1955 is one of the first nuclear sites in France. It combines the activities of the Research Centre of the French Atomic Energy Commission (CEA) and AREVA industrial operations. Today, a large part of the operations on this site consists of the cleaning and the dismantling of nuclear Installations, once the end of their life cycle has been reached. An example can be the reprocessing plant UP1. This unit, started in 1958 has been stopped in 1997 and its dismantling started quickly thereafter. Technical challenges of the UP1 dismantling are mainly linked to a very high risk of exposure due to a large variety of contaminated equipments and residuals of fission products, potential sources of irradiation. The dismantling of Hall 71 is a typical example of such challenge. This paper will present a solution developed by AREVA Clean-Up business unit, in collaboration with COFIM Industry, to remove contamination incrusted inside the pipes before starting the cutting operations, thus reducing irradiation risk. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Recovery from feeding unconditioned feed to part of the solvent extraction process in Thorp - 14430

A recent event in the Thorp Reprocessing Plant involved unconditioned feed being sent to one of the main process solvent extraction vessels. As the feed was unconditioned the fissile material was lost to the aqueous raffinate stream. The plant was tripped and shutdown in line with the baseline criticality safety case. This paper describes the recovery option that was taken to remove the Pu from the raffinate tank and safely consign it to the downstream effluent vessels. Key assumptions that were fundamental to maintaining criticality safety throughout these operations were clearly highlighted in the supporting criticality safety assessment. Key learning points that are of relevance to the wider criticality safety assessment community are pulled out at the end of this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Understanding Uranium Binding Mechanisms and Speciation on Stainless Steel - 17462

Stainless steels have been exposed to nitric acid solutions containing natural uranium under similar conditions to those found in reprocessing plants. Solution and solid analysis of the stainless steel samples has been undertaken to describe the contamination phenomenon. General corrosion of the surface was seen to take place over the month-long experiments. Contamination has been shown on all exposed surfaces. Solution analysis indicates that an equilibrium state is reached after 2 weeks exposure to the 12 M HNO{sub 3}. Surface analysis through autoradiography and X-ray absorption near edge structure (XANES) indicate that multiple mechanisms of uptake may occur and consequently concentrated regions of contamination may be produced. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2019

Revisit of Phase Separation of Borosilicate Glass for Vitrification of High Level Radioactive Waste - 19166

Phase separation of glasses for vitrification of high level radioactive waste has been experimentally studied to know uncertainty of the phase separation conditions; temperature and glass compositions as well as to propose possibility of recycling of heavy metals on a longer time scale. Our current results at 700 and 750 deg. C for ternary borosilicate glass system showed that the phase separation region at 700 deg. C spreads along with a direction of increasing Na{sub 2}O composition when compared with the data in literature. On the other hand, at the glass composition employed in the reprocessing plant in Japan, no segregation was observed. Concerning the application of phase separation phenomena, we had previously reported recovery of almost all elements from the vitrified borosilicate glass containing simulated fission products through phase separation of the borosilicate glass and this suggests that we could propose that after vitrification of high-level radioactive waste heavy metals may be retrieved as energy or materials resources when they are required to utilize on a longer time scale. From this standpoint, phase separation of lead borosilicate glass was also studied to seek economical and versatile recycling. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Study on release and transport of aerial radioactive materials in reprocessing plants

The release and transport characteristics of radioactive materials at a boiling accident of the high active liquid waste (HALW) in a reprocessing plant have been studied for improving experimental data of source terms of the boiling accident. In the study, a heating test and a thermogravimetry and differential thermal analysis (TG-DTA) test were conducted. In the heating test using a simulated HALW, it was found that ruthenium was mainly released into the air in the form of gas and that non-volatile elements were released into the air in the form of mist. In the TG-DTA test, the rate constants and reaction heat of thermal decomposition of ruthenium nitrosyl nitrate were obtained from TG and DTA curves. (authors)

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Capture of volatile RuO 4 from oxidized simulated used nuclear fuel solutions

Ruthenium is a challenging fission product in used nuclear fuel (UNF) reprocessing due to its complex redox chemistry, variable speciation in nitric acid, partial extractability, and volatility. This work presents a strategy for Ru removal based on the volatilization of RuO 4 using the strong oxidant sodium bismuthate, followed by RuO 2 deposition onto various substrates. Among the materials tested, polymer-based substrates such as polyolefin wax film (Parafilm®), exhibited superior performance, achieving up to near-quantitative Ru removal from solution. After dissolution of the substrate, 99.6% of the Ru was recovered as RuO 2 . The deposition mechanism onto the polyolefin wax film involves both physisorption and chemical reaction through oxidative cleavage of olefinic bonds within the polymer. In contrast, on inorganic substrates such as Al foil, RuO 4 undergoes incomplete surface adsorption and reduction, as confirmed by FTIR and XPS analyses. This approach remains effective under conditions of simulated oxidation of UNF solutions, selectively removing Ru while leaving other fission products in solution. As a result, capture of RuO 4 offers a practical and efficient strategy for ruthenium decontamination and recovery, offering a route for integration into existing UNF reprocessing flowsheets to enhance overall process safety and performance.

Ruthenium↗
DOE OSTI2015

Improvements to neutron detection of Pu in solvent for an aqueous buffer tank, magnox reprocessing plant, Sellafield - 14216

The PP Active Feed Tank (PPAFT) is a buffer tank that is not safe by geometry and which is located between the Medium Active (MA) and Plutonium Purification (PP) cycles of the chemical separation process in the Magnox Reprocessing Plant. The PPAFT receives Pu nitrate in acid which has been conditioned ready for the extraction of Pu with TBP/OK (tri-butyl phosphate in kerosene) solvent in the PP Cycle. The criticality safety case for the PPAFT recognises that during both normal operations and potential maloperations solvent which is dissolved (TBP which is present in solution but can be released upon a change in acidity) and entrained (TBP/OK physically carried over but not in solution) will accumulate in the PPAFT. The only practicable route out for Pu in solvent in the PPAFT is the normal process route out from the base of the PPAFT to the PP cycle through intermediate vacuum lift and feed vessels. However, the dished end base of the PPAFT and the intermediate vacuum lift and feed vessels are of particularly challenging geometry. There are five neutron monitors placed around the PPAFT for the detection of Pu in solvent. The original design intent of the neutron detection system was that any single monitor should be able to reveal an accumulation of Pu in solvent before it would potentially be unsafe if fed downstream. However, over time there have been changes to the fuel feed and chemical process of the plant which have had the potential to affect the performance of the neutron monitoring system. The ability of the neutron monitoring system to function to its original intent was questioned. Work has recently been carried out to investigate the capabilities of the PPAFT neutron monitoring system for current plant conditions and to identify potential improvements while making best use of the existing neutron monitors. It was found that moving a neutron monitor from the side to the top of the PPAFT and adding alarms based on the ratio of signals from multiple neutron detectors would be effective in improving detection of Pu in solvent. This paper describes the challenges involved in detecting thin layers of Pu in solvent using neutron monitors, discusses the improvements made to the PPAFT neutron monitors and details the combination of real plant data and modelling with the shielding code MCBEND used to provide confidence in the enhanced neutron monitoring system. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Capacity of ETS-10 supported carbon nanosorbent for removal of multicomponent from off-gas stream

Nuclear fission results in fission and activation products, some of which are volatile under the conditions of used nuclear fuel (UNF) reprocessing. Several volatile radionuclides are released into the atmosphere through offgas stream, which includes {sup 3}H, {sup 14}C, {sup 129}I and noble gases, Kr and Xe. {sup 129}I is a highly mobile and volatile contaminant. It has half-life period about 1.57 x 107 years that tends to bioaccumulate, concentrating in the thyroid glands and affecting bio-metabolism in humans. {sup 85}Kr is a chemically inert radioisotope with a half-life of 10.7 years that continuously accumulate in the environment. Control of emissions of these contaminants is necessary to comply with the stringent regulatory limits placed on their releases to protect human health and environment. Removal of these contaminants has been a challenge since the 1950's, and several methods including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Currently, adsorption on porous solid sorbent shows the most promise results. Up to now, several sorbents such as silver impregnated mordenite, silver ETS-10, chalcogen based aerogels, and metal organic framework, MOFs have been developed for this purpose. Despite these advances, development of economical and highly active sorbents for complete removal and immobilization of volatile radioactive gases remains a key challenge in realizing advanced fuel cycles. Present research work is focused on the determination of the removal capacity of C'at'ETS-10 sorbent for a multicomponent (iodine and krypton) off-gas stream at ambient temperature. Our previously reported work showed that the 10 wt% C'at'ETS-10 sorbent was active for adsorption of krypton as well as iodine at ambient temperature and pressure. Therefore, 10 wt% C'at'ETS-10 sorbent was synthesized by wet-impregnation technique. The performance evaluation of the prepared nanosorbent was examined for the removal of iodine and krypton from the breakthrough curves obtained using bench scale experimental setup. ETS-10 supported carbon nanoparticles, C'at'ETS-10 nanosorbent was successfully synthesized by wet-impregnation technique. The 1/8 inch pelletized sorbent was used to minimize the pressure drop in adsorption column. The capacity of sorbent was found to be 0.143 mg g{sup -1} for krypton and 24.5 mg g{sup -1} for iodine. This experimental results indicates the C'at'ETS-10 is the effect sorbent for removal of multicomponent, krypton as well as iodine from the off-gas stream at ambient temperature. (authors)

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Initial recovery of the magnox reprocessing plant following a loss of reductant feed in the solvent extraction process - 14429

A recent event in the Magnox Reprocessing Plant involved the loss of reductant feed to a contacting vessel and required the plant to be shutdown. The criticality team attended the site out-of-hours to facilitate the production of a recovery plan. This initial recovery plan involved manual dosing of the contactor with reductant but was heavily dependent upon initial information regarding the length of time the maloperation had been occurring. Key assumptions fundamental to maintaining criticality safety during the planned recovery were clearly highlighted in the assessment that was written to support the operation, which allowed informed decisions to be made regarding the adequacy of the plan, once further information emerged suggesting the fault had been ongoing for much longer than originally assumed and therefore the amount of Pu in the contactor was much higher than initially thought. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Design and Test Plan for an Integrated Iodine Scrubber and Polishing Bed System

The capture and subsequent immobilization of four regulated volatile radionuclides ( 3 H, 14 C, 85 Kr, and 129 I) and relevant semivolatile species from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of significant research interest on the part of the US Department of Energy and other international organizations. Significant research and development has been conducted over the past decade. In 2016 an initial engineering evaluation and design of the off-gas abatement systems required for a hypothetical 1000 t/yr UNF reprocessing facility treating 5 yr–cooled, 60 GWd/tIHM UNF was completed. One of the key findings of that report was that the consumption rate of silver-based iodine sorbents in the dissolver off-gas primary iodine capture bed is very high and may warrant the evaluation of alternative methods to capture the bulk of the iodine that could significantly reduce the associated frequent remote handing of the iodine filter beds. This report is intended to describe the design of an experimental system that can be used to examine the use of aqueous scrubbing to remove the bulk of the iodine from the dissolver off-gas stream prior to a silver-based solid sorbent that would be used to provide the final iodine capture or polishing step. This report also provides a description of the initial series of tests that are proposed for this system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗