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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Performance of Silver-Exchanged Mordenite and Silver-Functionalized Silica-Aerogel Under Vessel Off-gas Conditions

US regulations 40 CFR 61, 40 CFR 190, and 10 CFR 20 govern release limits for volatile radionuclides contained in the gaseous effluents of used nuclear fuel (UNF) reprocessing facilities. Of the four volatile radionuclides that are restricted by the release limits ( 3 H, 14 C, 85 Kr, and 129 I) 129 I will require the greatest degree of abatement. Previous studies have shown that overall plant decontamination factor (DF) for 129 I must, at minimum, exceed 1,000 to meet regulatory requirements. Iodine-containing off-gas will be present in the dissolver off-gas, the cell off-gas, the vessel off-gas (VOG), the waste off-gas, and the shear off-gas. The VOG will most likely contain 1%–5% of the total iodine at part per billion (ppb) concentrations. A number of studies have examined iodine abatement from the dissolver off-gas, which contains greater than 95% of the iodine inventory of the plant, but very few have examined the recovery of iodine from the VOG stream. Reduced silver-exchanged mordenite (Ag 0 Z) has been considered a promising iodine sorbent for use in UNF reprocessing plants, but recent efforts have identified silver-functionalized silica-aerogel (AgAerogel) as a potential alternative. The testing reported here reflects a substantial increase in the body of work available on iodine adsorption by both silver sorbents under prototypical VOG conditions. Both I2 and methyl iodide (CH 3 I) adsorption tests have been conducted with Ag 0 Z and AgAerogel at a range of concentrations up to 1,000 ppb. These experiments were performed with two specific aims: (1) to assess the effect of iodine concentration on its adsorption by the sorbent and (2) to compare the performance of Ag 0 Z and AgAerogel in the removal of iodine from a prototypical VOG stream.

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

CoDCon Project (Final Report)

The co-decontamination (CoDCon) project was established in FY 2016 with the objectives of (a) evaluating the uncertainty in the uranium (U)/plutonium (Pu) ratio in a mixed U/Pu product from a tributyl phosphate (TBP)–based solvent extraction flowsheet, and (b) developing and demonstrating on-line optical spectroscopy for real-time monitoring of key components (e.g., Pu, U, and HNO 3 concentrations) in the process solutions. We were interested in assessing the accuracy and precision to which a specific uranium-to-plutonium (U/Pu) ratio can be achieved, which for the purposes of this project was set at a U/Pu mass ratio of 7/3. The uncertainty associated with achieving this specific target U/Pu ratio was investigated during five flowsheet tests using laboratory-scale solvent extraction equipment. In addition, optical spectroscopic techniques were incorporated into the CoDCon solvent extraction testing system, allowing real time monitoring of all input and output process streams. Two CoDCon flowsheet tests were performed in FY 2018 using a simple dissolved fuel simulant containing only U (~1 M) and Pu (~15 mM) in nitric acid (HNO 3 ; ~3 M). In FY 2019, two additional flowsheet tests were performed. For the first of these (CoDCon Run 3), the dissolved fuel simulant was similar to that used in the first two tests, with the inclusion of 1 mM neptunium (Np). The second test conducted in FY 2019 (CoDCon Run 4) used a more representative dissolved fuel simulant, including addition of non-radioactive fission product elements. A fifth CoDCon flowsheet test (CoDCon Run 5) was conducted in FY 2020, with the following additional objectives: (1) routing of the technetium (Tc) in the simulated dissolved fuel solution to the solvent extraction raffinate, and (2) routing of the Np in the simulated dissolved fuel solution to the U/Pu product. All tests used a bank of sixteen 2 cm centrifugal contactors. The tests involved first loading the solvent (30 vol% TBP dissolved in n-dodecane) with U and Pu (and Np, for Run 5), then the Pu (and Np) was stripped from the loaded solvent with a U(IV) solution (~50 mM) and the flowsheet conditions were adjusted such that some U partitioned into the Pu-containing product stream. The amount of U accompanying the Pu was monitored in real time using optical spectroscopic techniques coupled with chemometric modeling. Based on the real-time spectroscopic measurement of the U/Pu ratio, adjustments were made to the flowrate of the fresh TBP solvent phase used to scrub U from the aqueous Pu-containing product. This proved to be a very effective way to control the U/Pu mass ratio in the product. This report presents the results of the CoDCon Run 5 test. The flowsheet tested in Run 5 was substantially different than that run in the prior tests, especially the solvent loading section of the flowsheet. Two key changes were made. First, based on the objective to extract all the Np and route it with the U/Pu product, pentavalent vanadium [V(V)] was added to the feed and scrub solutions. The purpose of the V(V) was to convert all the Np to the +6 oxidation state, which is extractable by TBP. Second, a high acid (8 M HNO 3 ) scrub was added to the flowsheet to scrub the Tc from the solvent. This was followed by a low acid scrub (0.05 M HNO 3 ) to reduce the residual HNO 3 concentration in the solvent prior to the Pu stripping step. The output from the low acid scrub was collected separately, rather than routing towards the raffinate. The modifications to the solvent loading part of the flowsheet were only partially successful. The treatment with V(V) was effective at converting the Np to Np(VI). Only 1.3% of the Np remained in the raffinate solution. However, ~40% of the Np stripped out of the solvent in the low acid scrub step; nearly 20% of the Pu also was stripped from the solvent during the low acid scrub. For further development, either modifications to the flowsheet, or concentration and recycle of the low acid stream into the

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Plan for Developing TRISO Fuel Processing Technologies

This Plan demonstrates the availability of technologies for processing TRISO used nuclear fuel for waste management and actinide recovery purposes. These technologies are judged to be at a very low level of technology readiness and as such they constitute a fertile research area for the DOE-NE’s Office of Materials and Chemical Technologies. Strategies to mature the technologies to a point where they can reasonably be considered in engineering alternatives analyses typically involve laboratory-scale tests using fuel simulant to characterize process streams and demonstrate key engineering features. Several criteria are available to help selecting candidate technologies for further maturation

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

HZ-PAN and AgZ-PAN Desorption Characterization

While previous research focused on adsorption of Kr and Xe on HZ-PAN and AgZ-PAN, recent focus has shifted to desorption. Preliminary testing was completed last year, providing valuable information to guild further research. Preliminary studies indicated that thermal swing operations, along with a sweep gas, are required for effective desorption from both AgZ-PAN and HZ-PAN. This series of tests focused on characterizing the bulk gases removed from each sorbent and relative separation of Kr and Xe from each other and from the bulk air. Single stage adsorption/ resulted in 99-100% separation of Xe from Kr, and 77-83% separation of Kr from air while concentrating Xe seven to tenfold and Kr three to fivefold when Xe is not present. Previous isothermal studies of both sorbents have shown that sorbent capacity is proportional to feed gas concentration. Therefore, it is anticipated that additional HZ-PAN stages will further concentrate and purify the final Kr stream with higher efficiency. It is expected that Xe will be released to atmosphere after desorption. However, it could also be purified further with more AgZ-PAN stages. Further testing should be done to determine maximum purity achievable for both Kr and Xe. Additionally, renewed focus on increasing Xe capacity on AgZ-PAN and Kr capacity on HZ-PAN is recommended.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Organic Iodine Capture from Vessel Off-gas

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 non-radioactive synthetic gas mixtures blended from air, nitrogen, NO, NO2, water, and the target iodine compounds diatomic iodine (I2), representing likely inorganic iodides, and methyl iodide (iodomethane, CH3I) and 1-iodobutane (butyl iodide) a surrogates for potential organic iodides, that could exist in dissolver off-gas (DOG) and vessel off-gas (VOG) streams from used nuclear fuel aqueous reprocessing.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2021

Effects of HZ-PAN Sorbent Irradiation on Kr Adsorption

Idaho National Laboratory (INL) has developed and studied an engineered form hydrogen mordenite in polyacrylonitrile matrix (HZ-PAN) for krypton (Kr) adsorption. Adsorption studies to date have successfully separated and captured Kr from carrier gas streams, but those studies have used non-radioactive gases for testing. However, if utilized in a used nuclear fuel (UNF) reprocessing facility, this sorbent will be exposed to the potential deleterious effects of radiation. This report explores Kr adsorption behavior on HZ PAN after the sorbent was exposed to gamma radiation. A batch of HZ-PAN was divided into five segments, a control segment that was not exposed to radiation, and the other four segments that were exposed to progressively increased levels of radiation dose, up to 5 MGy. Samples were then analyzed to determine total surface area as well as their capacity to capture Kr. This study reveals HZ PAN retains its initial surface area and Kr-capture capacity after radiation exposure, indicating viability for service in radioactive gases.

42 - ENGINEERING↗
DOE OSTI2022

Krypton Concentration using HZ-PAN

Idaho National Laboratory (INL) has developed and tested engineered sorbents to separate and capture volatile fission products such iodine, xenon (Xe), and krypton (Kr) from off-gas streams. As noble gases, Xe and Kr can be difficult to capture and separate. Historically, cryogenic distillation has been used to execute the separation. However, performing this separation from bulk air streams is expensive and can pose significant hazards. INL has successfully developed two different sorbents for Xe and Kr capture, silver mordenite polyacrylonitrile and hydrogen mordenite polyacrylonitrile (HZ-PAN). Adsorption studies to date successfully separated and captured Kr from carrier gas streams, but those studies focused primarily on initial separation and capture. Successful utilization in a used nuclear fuel (UNF) reprocessing facility, however, requires further concentration of the Kr to minimize long term storage volumes. This study focuses on the concentration of Kr utilizing HZ PAN as the concentrating media. Concentrating Kr during desorption is ideal for two reasons: one, to produce Kr that can be used for commercial and research applications, or two, to minimize the volume of the radioactive gaseous waste stream for disposal. This report investigates the Kr concentrating potential of HZ-PAN by running multiple Kr adsorption-desorption cycles, collecting the desorbed effluent from a saturated HZ-PAN column, and loading it onto a fresh column. This study demonstrates that four adsorption-desorption cycles can transform a 150 ppmv Kr stream into a 37.5% Kr stream, an overall concentration factor of 2497. Each concentration step results in successively smaller volumes desorbed, reducing volume by a factor of approximately 40,000. Depending on the operational goals, during desorption one could collect smaller volume fractions of high concentration Kr (~ 67%). This demonstration should be considered proof-of-concept. Further refinement is necessary to develop optimum operating schemes to integrate into UNF off-gas treatment.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2016

Thermal Stress in HFEF Hot Cell Windows Due to an In-Cell Metal Fire

This work investigates an accident during the pyrochemical extraction of Uranium and Plutonium from PWR spent fuel in an argon atmosphere hot cell. In the accident, the heavy metals (U and Pu) being extracted are accidentally exposed to air from a leaky instrument penetration which goes through the cell walls. The extracted pin size pieces of U and Pu metal readily burn when exposed to air. Technicians perform the electrochemical extraction using manipulators through a 4 foot thick hot cell concrete wall which protects them from the radioactivity of the spent fuel. Four foot thick windows placed in the wall allow the technicians to visually control the manipulators. These windows would be exposed to the heat of the metal fire. As a result, this analysis determines if the thermal stress caused by the fire would crack the windows and if the heat would degrade the window seals allowing radioactivity to escape from the cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

"Ask Argonne" - Dave Grabaskas, Nuclear Engineer, Part 2

Part 1 (http://www.youtube.com/watch?v=Vs_0wXoSL8M) of Dr. Dave Grabaskas' "Ask Argonne" video set drew many questions from the public. In Part 2, Grabaskas answers three of those questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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

Electrochemical Manipulation and Radiolytic Evaluation of Organic Phase Neptunium

Presentation for Seed LDRD proposal presentation: This Seed LDRD research proposes to test an innovative approach to precisely control the oxidation state distribution of neptunium (Np) under used nuclear fuel (UNF) reprocessing conditions using novel electrochemical methods. Current UNF reprocessing technologies are not financially viable in the US, requiring significant scientific and technological innovation to improve cost efficiency. Reducing the number of process cycles is one avenue for increasing cost efficiency, of which resolving the challenge associated with isolating Np is critical. Under envisioned process conditions, Np is present in a mixture of extractable (Np(IV) and Np(VI)) and inextractable (Np(V)) oxidation states, the distribution of which is dependent on several factors, including ionizing radiation dose, that change throughout a reprocessing scheme. Consequently, Np unintentionally partitions into various UNF reprocessing phases and product streams, ultimately increasing the number of process cycles to isolate UNF components. Here, we propose to employ novel, high surface area, optically transparent, Ligand Modified tin-doped indium oxide Electrodes (LMES). These LMEs bind Np, and therefore enable concurrent generation and spectroscopic characterization of Np oxidation states in organic solutions. These proof-of-concept experiments will: (i) facilitate optimization of the proposed electrochemical system conditions for Np oxidation state manipulation; (ii) identify the accessible electrochemical window for Np complexes in the organic phase; (iii) report characteristic optical spectra for each accessible complexed Np oxidation state; and (iv) determine the lifetime and partitioning of a given atypical Np oxidation state in the organic phase in the presence and absence of ionizing radiation fields. The data gathered by this Seed LDRD will provide support for the design of an electrochemical process concept for the precise manipulation of Np oxidation states in UNF reprocessing solvent systems, with the intention of providing advanced control over Np mass transfer, and thus greater process efficiency and economy.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Radiolytic transformation of AHA under single-cycle conditions

Historically, the radiation robustness of extractant ligands and the impact of their radiolysis products have been used to evaluate the radiolytic feasibility of solvent system formulations for used nuclear fuel (UNF) reprocessing. However, other key additives (e.g., reducing and hold-back reagents) are employed by these solvent systems. These chemical species have typically been overlooked from a radiolytic perspective, and consequently little is known on how they impact process performance. Therefore, to holistically support the development of advanced, simplified, single-cycle flowsheets for UNF reprocessing, the radiation robustness of two key proposed additives – (i) the complexing reductant acetohydroxamic acid (AHA) and (ii) the hold-back reagent 1,2-cyclohexylenedinitrilotetraacetic acid (CDTA) – have been evaluated under envisioned process conditions using a combination of steady-state gamma and time-resolved pulsed electron irradiation techniques.

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 OSTI2013

Polyvalent fuel treatment facility (TCP): shearing and dissolution of used fuel at La Hague facility

Although many used nuclear fuel types have already been recycled, recycling plants are generally optimized for Light Water Reactor (LWR) UO{sub x} fuel. Benefits of used fuel recycling are consequently restricted to those fuels, with only limited capacity for the others like LWR MOX, Fast Reactor (FR) MOX or Research and Test Reactor (RTR) fuel. In order to recycle diverse fuel types, an innovative and polyvalent shearing and dissolving cell is planned to be put in operation in about 10 years at AREVA's La Hague recycling plant. This installation, called TCP (French acronym for polyvalent fuel treatment) will benefit from AREVA's industrial feedback, while taking part in the next steps towards a fast reactor fuel cycle development using innovative treatment solutions. Feasibility studies and R/Development trials on dissolution and shearing are currently ongoing. This new installation will allow AREVA to propose new services to its customers, in particular in term of MOX fuel, Research Test Reactors fuel and Fast Reactor fuel treatment. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Savannah River Site HB-Line Plutonium Oxide Production Process Simulation Model - 17458

The AFS II/Plutonium Oxide Production process is in its initial operational phase at the Savannah River Site's (SRS) HB-Line. The process is new for the facility where it is being performed, but the facility itself has been in use on and off for over 50 years. Beginning this process inside the facility has resulted in multiple process delays, due to both equipment issues and process restrictions. In order to better understand the process and to identify ways to improve facility performance, Savannah River Nuclear Solutions (SRNS), in conjuncture with Newport News Shipbuilding (NNS), developed the HB-Line Process Simulation Model. This tool completely replicates the operations that comprise the HB-Line process and simulates all of the constraints associated with the process. The model realistically accounts for the number of personnel required for a process step, the time required for a process step to occur, rules related to the parallel operation of process steps, and realistic equipment and operational delay data. The result is a tool that allows SRNS to test different scenarios for improving the process throughput, while realistically accounting for potential process delays and restrictions. The tool is currently in use and has allowed SRNS to present to their customer, the National Nuclear Security Administration (NNSA), estimated production capabilities over wide-ranging timetables and assumptions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗