Spent caustic liquors (SCL) generated from crude oil refineries have characteristics of high COD (chemical oxygen demand) contents and relatively small generation rates as compared with general wastewater ones. The odorous naphthenates, phenolates, and sulfides in the liquors adversely affect the normal operation of the related wastewater treatment plants and effluent water qualities. This study aims at the reduction of COD in a naphthenic spent caustic liquor generated from a domestic refinery with crude processing and naphtha cracking units. Primary tests indicated that around 50% of the COD in the SCL could be biodegraded. Chemical oxidation methods were tried to possibly upgrade the COD removal. Experiments indicated that acidification of the SCL sample to pH 2-3 could reduce the COD from an average of 51,600 to 20,800 mg/L by removing the separated naphthenic acids. Fenton¡¦s method with oxidants of 20 mL/L 35% H2O2 and FeSO4.7H2O 5 g/L, oxidation time of 1 hour at conditions of pH 2-3 and 80-100oC, could reduce the COD of the acidified SCL from an average of 20,800 to 11,100 mg/L. The overall COD removal was around 78% and the efficiency is comparable to that of a traditional Wet Air Oxidation (WAO) process of around 75%. Economic analysis indicated that for treating the SCL of 80 m3/day by the traditional WAO process, an initial equipment cost of 10 millions USD and annual operating cost of around 1.5 millions USD are required. By the developed acidification-Fenton¡¦s process, an initial equipment cost of 0.7 million USD and annual operating cost of around 0.5 million USD are expected. The developed process can be superior to the WAO one.

Reprocessing nuclear fuel is becoming more viable in the United States due to the anticipated increase in construction of nuclear power plants, the growing stockpile of existing used nuclear fuel, and a public desire to reduce the amount of this fuel. However, a new reprocessing facility in non-weapon states must be safeguarded and new reprocessing facilities in weapon states will likely have safeguards due to political and material accountancy reasons. These facilities will have state of the art controls and monitoring methods to safeguard special nuclear materials, as well as to provide real-time monitoring. The focus of this project is to enable the development of a safeguards strategy that uses well established photon measurement methods to characterize samples from the UREX+3a reprocessing method using a variety of detector types and measurement times. It was determined that the errors from quantitative measurements were too large for traditional safeguards methods; however, a safeguards strategy based on qualitative gamma ray and neutron measurements is proposed. The gamma ray detection equipment used in the safeguard strategy could also be used to improve the real-time process monitoring in a yet-to-be built facility. A facility that had real-time gamma detection equipment could improve product quality control and provide additional benefits, such as waste volume reduction. In addition to the spectral analyses, it was determined by Monte Carlo N Particle (MCNP) simulations that there is no noticeable self shielding for internal pipe diameters less than 2 inches, indicating that no self shielding correction factors are needed. Further, it was determined that HPGe N-type detectors would be suitable for a neutron radiation environment. Finally, the gamma ray spectra for the measured samples were simulated using MCNP and then the model was extended to predict the responses from an actual reprocessing scenario from UREX+3a applied to fuel that had a decay time of three years. The 3-year decayed fuel was more representative of commercially reprocessed fuel than the acquired UREX+3a samples. This research found that the safeguards approach proposed in this paper would be best suited as an addition to existing safeguard strategies. Real-time gamma ray detection for process monitoring would be beneficial to a reprocessing facility and could be done with commercially available detectors.

In this study, various techniques of hydrometallurgy and ultrasound-assisted solvent extraction were used to recover the platinum group metals (PGM¡¦s) from a composite sample of honeycomb-type autocatalysts. After they were removed from the converter casings, the autocatalyst substrates were first crushed and then ground by a ball mill. The recovery procedures employed are shown as follows: (1) dissolve PGM¡¦s from ground spent autocatalyst by aqua regia leaching; (2) separate PGM¡¦s from base metals in the aqua regia leachate by metal cementation using zinc powder so that PGM¡¦s can be precipitated out; (3) the PGM¡¦s precipitate was first dissolved by aqua regia, then proceed to remove nitrate and hydrochloride within. The residue was further dissolved in hydrochloride acid as a preparation step for solvent extraction; (4) the PGM¡¦s pregnant solution of hydrochloride acid was treated by solvent extraction and stripping to separate and purify each component of PGM¡¦s. Effects of ultrasound agitation on the efficiency of solvent extraction was also evaluated in this work. Results of aqua regia leaching experiments have shown that the quantity of dissolved PGM¡¦s increased as the solid-to-liquid ratio decreased. The maximum dissolved quantity of PGM¡¦s could be obtained by a 3-hr leaching time. At this stage, the PGM¡¦s recoveries are 80-90% for platinum and rhodium and greater than 99% for palladium. The result of a preliminary test has indicated that acetic acid can not effectively separate the PGM¡¦s and base metals. Thus, the method of cementation by zinc powder was employed to separate PGM¡¦s from base metals. Before cementation, the aqua regia leachate was diluted and pH-adjusted to greater than 2. In so doing, an almost complete cementation (>99%) could be obtained by the least quantity of zinc powder. In addition, the base metals occurred with the PGM¡¦s precipitate have been minimized except lead and zinc. While palladium was extracted by di-n-octyl sulfide (DOS), ultrasound assistance has rendered a complete extraction within a few minutes. At this stage, the extraction efficiency was found to be independent of the HCl concentration. It was found that platinum and rhodium were not extracted by DOS. When platinum was extracted by tri-n-octylamine (TOA) and assisted by ultrasound, rhodium will be extracted at the HCl concentration higher than 4M. Thus, TOA is not an effective chemical for selective extraction of platinum. TOA was then replaced by tributyl phosphate (TBP). Experimental results have indicated that the extraction of platinum using TBP was affected by the HCl concentration. The best result was obtained when the HCl concentration was 5M. Extraction by TBP was found to be fast. It took only 20-30 seconds to reach the equilibrium even with no ultrasound assistance. But multi-stage extractions are generally required to extract platinum completely. Rhodium was found to be not extracted by TBP. After palladium and platinum were extracted, only rhodium was remained in the reffinate. In summary, solvent extraction using DOS and TBP has made it possible to separate palladium, platinum, and rhodium effectively. In the palladium stripping solution almost no base metals was determined. However, zinc and lead were found in the platinum stripping solution and the rhodium-containing raffinate. These base metals should be removed to achieve a better purity for each precious metal. The TCLP (i.e., a leaching test for toxicity) result of the autocatalyst substrate after aqua regia leaching has found to be non-hazardous. However, several streams of wastewater and acid gas generated in the recovery process should be properly managed to avoid the secondary pollution.

In the event that a radiological dispersal device (RDD) is detonated in the U.S. or near U.S. interests overseas, it will be crucial that the actors involved in the event can be identified quickly. If irradiated nuclear fuel is used as the dispersion material for the RDD, it will be beneficial for law enforcement officials to quickly identify where the irradiated nuclear fuel originated. One signature which may lead to the identification of the spent fuel origin is the isotopic composition of the RDD debris. The objective of this research was to benchmark a forward model methodology for predicting isotopic composition of spent nuclear fuel used in an RDD while at the same time optimizing the fidelity of the model to reduce computational time. The code used in this study was Monteburns-2.0. Monteburns is a Monte Carlo based neutronic code utilizing both MCNP and ORIGEN. The size of the burnup step used in Monteburns was tested and found to converge at a value of 3,000 MWd/MTU per step. To ensure a conservative answer, 2,500 MWd/MTU per step was used for the benchmarking process. The model fidelity ranged from the following: 2-dimensional pin cell, multiple radial-region pin cell, modified pin cell, 2D assembly, and 3D assembly. The results showed that while the multi-region pin cell gave the highest level of accuracy, the difference in uncertainty between it and the 2D pin cell (0.07% for 235U) did not warrant the additional computational time required. The computational time for the multiple radial-region pin cell was 7 times that of the 2D pin cell. For this reason, the 2D pin cell was used to benchmark the isotopics with data from other reactors. The reactors from which the methodology was benchmarked were Calvert Cliffs Unit #1, Takahama Unit #3, and Trino Vercelles. Calvert Cliffs is a pressurized water reactor (PWR) using Combustion Engineering 14??14 assemblies. Takahama is a PWR using Mitsubishi Heavy Industries 17??17 assemblies. Trino Vercelles is a PWR using non-standard lattice assemblies. The measured isotopic concentrations from all three of the reactors showed good agreement with the calculated values.

A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.

<p>The kinetics for radiation induced dissolution of spent nuclear fuel is a key issue in the safety assessment of a future deep repository. Spent nuclear fuel mainly consists of UO₂ and therefore the release of radionuclides (fission products and actinides) is assumed to be governed by the oxidation and subsequent dissolution of the UO₂ matrix. The process is influenced by the dose rate in the surrounding groundwater (a function of fuel age and burn up) and on the groundwater composition. In this licentiate thesis the effects of HCO₃- (a strong complexing agent for UO2²⁺) and ionic strength on the kinetics of UO₂ oxidation and dissolution of oxidized UO₂ have been studied experimentally.</p><p>The experiments were performed using aqueous UO₂particle suspensions where the oxidant concentration was monitored as a function of reaction time. These reaction systems frequently display first order kinetics. Second order rate constants were obtained by varying the solid UO₂surface area to solution volume ratio and plotting the resulting pseudo first order rate constants against the surface area to solution volume ratio. The oxidants used were H₂O₂(the most important oxidant under deep repository conditions), MnO₄- and IrCl₆^2-. The kinetics was studied as a function of HCO₃- concentration and ionic strength (using NaCl and Na₂SO₄as electrolytes).</p><p>The rate constant for the reaction between H₂O₂ and UO₂ was found to increase linearly with the HCO3- concentration in the range 0-1 mM. Above 1 mM the rate constant is independent of the HCO3- concentration. The HCO₃- concentration independent rate constant is interpreted as being the true rate constant for oxidation of UO₂ by H₂O₂ [(4.4 ± 0.3) x 10-6 m min-1] while the HCO3- concentration dependent rate constant is used to estimate the rate constant for HCO₃- facilitated dissolution of UO₂2+ (oxidized UO₂) [(8.8 ± 0.5) x 10-3 m min-1]. From experiments performed in suspensions free from HCO₃- the rate constant for dissolution of UO₂2+ was also determined [(7 ± 1) x 10^-8mol m^-2 s^-1]. These rate constants are of significant importance for simulation of spent nuclear fuel dissolution.</p><p>The rate constant for the oxidation of UO₂ by H₂O₂ (the HCO₃- concentration independent rate constant) was found to be independent of ionic strength. However, the rate constant for dissolution of oxidized UO₂ displayed ionic strength dependence, namely it increases with increasing ionic strength.</p><p>The HCO₃- concentration and ionic strength dependence for the anionic oxidants is more complex since also the electron transfer process is expected to be ionic strength dependent. Furthermore, the kinetics for the anionic oxidants is more pH sensitive. For both MnO₄- and IrCl₆2- the rate constant for the reaction with UO₂was found to be diffusion controlled at higher HCO3- concentrations (~0.2 M). Both oxidants also displayed ionic strength dependence even though the HCO₃- independent reaction could not be studied exclusively.</p><p>Based on changes in reaction order from first to zeroth order kinetics (which occurs when the UO₂ surface is completely oxidized) in HCO₃- deficient systems the oxidation site density of the UO₂ powder was determined. H₂O₂and IrCl₆2- were used in these experiments giving similar results [(2.1 ± 0.1) x 10-4 and (2.7 ± 0.5) x 10^-4 mol m^-2, respectively].</p>

<p>Denna licensiatavhandling behandlar oxidativ upplösning av UO2. Upplösning av UO2 studeras huvudsakligen då UO2-matrisen hos använt kärnbränsle förväntas fungera som en barriär mot frigörande av radionuklider i ett framtida djupförvar. Lösligheten av U(IV) är mycket låg under i djupförvaret rådande förhållanden emedan U(VI) har betydligt högre löslighet. Oxidation av UO2-matrisen kommer därför att påverka dess löslighet och därmed dess funktion som barriär. I denna avhandling studeras den relativa effektiviteten av en- och två-elektronoxidanter för upplösning av UO2. Vid låga oxidantkoncentrationer är utbytet för upplösningen för en-elektronoxidanter signifikant lägre än för två-elektronoxidanter. För en-elektronoxidanter ökar dock utbytet med ökande oxidanthalt, vilket kan förklaras av den ökade sannolikheten för två konsekutiva en-elektronoxidationer av samma reaktionssite och den ökade möjligheten till disproportionering.</p><p>Radikaler och molekylära radiolysprodukters relativa inverkan på oxidativ upplösning av UO2 studeras också i denna avhandling genom mätning av mängden upplöst U(VI) i γ-bestrålade system som dominerades av olika oxidanter. Dessa studier visade att upplösningshastigheten av UO2 kan uppskattas från oxidantkoncentrationer framtagna genom simuleringar av radiolys i motsvarande homogena system och hastighetskonstanterna för ytreaktionerna. Simuleringarna visar att de molekylära oxidanterna kommer vara de viktigaste oxidanterna i alla system i denna studie vid långa bestrålningstider (>10 timmar). Vid liknande simuleringar av α-bestrålade system fanns att vid förhållanden relevanta för ett djupförvar för använt kärnbränsle, är det endast de molekylära oxidanterna (i huvudsak H2O2) som är av betydelse för upplösningen av bränslematrisen.</p><p>Då använt kärnbränsle innehåller en mängd radionuklider som utsätter UO2-matrisen för kontinuerlig bestrålning, är det av vikt att undersöka hur bestrålning påverkar reaktiviteten av UO2. Bestrålningseffekten på reaktionen mellan UO2 och MnO4- studerades. Dessa försök visade att bestrålning av UO2 vid doser >40 kGy leder till att reaktiviteten ökar upp till 1.3 gånger reaktiviteten av obestrålad UO2. Den ökade reaktiviteten kvarstår efter bestrålningen och effekten kan därför möjligen tillskrivas permanenta förändringar i materialet. Vid uppskattning av reaktiviteten hos använt kärnbränsle måste hänsyn tas till denna effekt då bränslet redan efter ett par dagar i reaktor blivit utsatt för doser >40 kGy.</p><p>Det har tidigare föreslagits att hastigheten för en heterogen västka/fast-fas reaktion är beroende av partikelstorleken hos det fasta materialet, vilket har studerats för UO2-partiklar i denna avhandling. Experimentellt bestämda kinetiska parametrar jämförs med de föreslagna ekvationerna för fyra storleksfraktioner av UO2-pulver och en UO2-pellet. Studien visade partikelstorleksberoendet av andra ordningens hastighetskonstant och aktiveringsenergin för oxidation av UO2 med MnO4- beskrivs relativt väl av de föreslagna ekvationerna.</p> / <p>The general subject of this thesis is oxidative dissolution of UO₂. The dissolution of UO₂ is mainly investigated because of the importance of the UO₂ matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO₂-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one- and two electron oxidants in dissolving UO₂is studied. The oxidative dissolution yield of UO₂was found to differ between one- and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation.</p><p>Furthermore, the relative impact of radical and molecular radiolysis products on oxidative dissolution of UO₂is investigated. Experiments were performed where the amount of dissolved U(VI) was measured in γ-irradiated systems dominated by different oxidants. We have found that the UO₂dissolution rate in systems exposed to γ-irradiation can be estimated from oxidant concentrations derived from simulations of radiolysis in the corresponding homogeneous systems and rate constants for the surface reactions. These simulations show that for all systems studied in this work, the molecular oxidants will be the most important oxidants for long irradiation times (>10 hours). Similar simulations of α-irradiated systems show that in systems relevant for a deep repository for spent nuclear fuel, only the molecular oxidants (mainly H₂O₂) are of importance for the dissolution of the fuel matrix.</p><p>The effect on UO₂ reactivity by irradiation of the material is of importance when predicting the spent fuel dissolution rate since the fuel, due to its content of radionuclides, is exposed to continuous self-irradiation. The effect of irradiation on the reaction between solid UO₂and MnO₄^- in aqueous solutions was studied. It was found that irradiation of UO2 at doses >40 kGy increases the reactivity of the material up to ~1.3 times the reactivity of unirradiated UO₂. The increased reactivity remains after the irradiation and can possibly be attributed to permanent changes in the material. This issue must be taken into account when predicting the reactivity of spent nuclear fuel since the fuel is exposed to doses >40 kGy after only a few days in the reactor.</p><p>It has earlier been suggested that the rate of a heterogeneous liquid-solid reaction depends on the size of the solid particles. This was investigated for UO₂particles in this thesis. Experimental kinetic parameters are compared to the previously proposed equations for UO₂ powder of four size fractions and a UO₂ pellet. We have found that the particle size dependence of the second order rate constant and activation energy for oxidation of UO₂ by MnO₄^- is described quite well by the proposed equations.</p>

Superheated steam (SS) drying of distillers’ spent grain (DSG) is a more energy efficient alternative to conventional hot air drying. SS drying at sub-atmospheric pressure (also referred to as low pressure) can prevent burning and lowering the quality of the food product. The objective of this study was to design, fabricate, and test a SS drying system that could operate at sub-atmospheric pressure for drying DSG. After the custom designed system was constructed, major problems associated with the system were identified. A number of tests were carried out and modifications were made to the system to resolve technical problems. Distillers’ spent grain was then successfully dried using the system under various levels of temperature from 95 to 115°C and pressure of either -25 or -20 kPa, with a SS velocity from 0.100 to 0.289 m/s.

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