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Design Optimization of A generic Fissile Solution for Mo99 production using Electron Beam-based Neutron Generator using MCNP+CFD

Molybdenum-99 ( 99 Mo) is a medical isotope used in 80% of all medical imaging procedures today. However, 99 Mo is not directly administered to patients for imaging; its decay product, technetium-99m ( 99 mTc), is a metastable isotope with a half-life of ~6 hours. Such a short half-life does not allow for production, separation, and application to a patient before decaying further. Therefore, the solution is to produce the precursor to 99 mTc, 99 Mo. 99 Mo has a half-life of 66 hours, long enough to produce the material and send it to the desired location before it fully decays. Under National Nuclear Security Administration’s (NNSA) highly enriched uranium (HEU) minimization mission, the Material Management and Minimization (M3) office leads the Molybdenum-99 (Mo-99) program. A major goal of the Mo-99 program is to develop methods of producing 99 Mo using low enriched uranium (LEU) opposed to the traditionally used HEU domestically in the U.S. Internationally, LEU is the new standard for use in research reactors and isotope production facility. In this study, we introduce a generic fissile solution system for 99 Mo production facility meeting the design requirement of 1) LEU usage as fissile material and 2) subcritical fissile system coupled with a horizontal electron beam (E-beam) accelerator. The focus of this research is to optimize a fissile solution configuration for maximum 99 Mo production yield using MCNP model, and ultimately implement the optimized design with computational fluid dynamics (CFD) model to understand thermal hydraulic behavior and solution convection characteristic. The report consists of two parts: 1) isotope yield calculation and criticality analysis with a wide range of parametric test matrix, and 2) thermal hydraulic analysis for solution mixing and cooling assessment.

60 APPLIED LIFE SCIENCES↗

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗

Evaluation of a measurement system for Uranium electrodeposition control to radiopharmaceuticals production

For 2016, studies by international bodies forecast a crisis in the supply of Molybdenum ({sup 99}Mo), which is the generator of {sup 99m}Tc, widely used for medical diagnoses and treatments. As a result, many countries are making efforts to prevent this crisis. Brazil is developing the Brazilian Multipurpose Reactor (RMB) project, under the responsibility of the National Nuclear Energy Commission (CNEN). The RMB is a nuclear reactor for research and production of radioisotopes used in the production of radiopharmaceuticals and radioactive sources, broadly used in industrial and research areas in Brazil. Electrodeposition of uranium is a common practice to create samples for alpha spectrometry and this methodology may be an alternative way to produce targets of low enriched uranium (LEU) to fabricate radiopharmaceuticals, as {sup 99}Mo, used for cancer diagnosis. To study the electrodeposition, a solution of 10 mM uranyl nitrate, in 2-propanol, containing uranium enriched to 2.4% in {sup 235}U, with pH = 1, was prepared and measurements with an alpha spectrometer were performed. These studies are justified by the need to produce {sup 99}Mo since, despite using molybdenum in bulk, Brazil is totally dependent on its import. In this project, we intend to obtain a process that may be technologically feasible to control the radiation targets for {sup 99}Mo production. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

Production of medical {sup 99m}Tc isotope via photonuclear reaction

{sup 99m}Tc with a 6 hour half-life is one of the most important medical isotopes used for the Single-Photon Emission Computed Tomography (SPECT) inspection in hospitals of US, Canada, Europe and Japan. {sup 99m}Tc isotopes are extracted by the milking method from parent {sup 99}Mo isotopes with a 66 hour half-life. The supply of {sup 99}Mo isotopes now encounters a serious crisis. Hospitals may not suitably receive {sup 99}Mo medical isotopes in near future, due to difficulties in production by research nuclear reactors. Many countries are now looking for alternative ways to generate {sup 99}Mo isotopes other than those with research reactors. We discuss a sustained availability of {sup 99m}Tc isotopes via the {sup nat}Mo(γ, n) photonuclear reaction, and discuss to solve technical problems for extracting pure {sup 99m}Tc isotopes from other output materials of photonuclear reactions.

07 ISOTOPE AND RADIATION SOURCES↗

Design and experimental activities supporting commercial U.S. electron accelerator production of Mo-99

{sup 99m}Tc, the daughter isotope of {sup 99}Mo, is the most commonly used radioisotope for nuclear medicine in the United States. Under the direction of the National Nuclear Security Administration (NNSA), Los Alamos National Laboratory (LANL) and Argonne National Laboratory (ANL) are partnering with North Star Medical Technologies to demonstrate the viability of large-scale {sup 99}Mo production using electron accelerators. In this process, {sup 99}Mo is produced in an enriched {sup 100}Mo target through the {sup 100}Mo({gamma},n){sup 99}Mo reaction. Five experiments have been performed to date at ANL to demonstrate this process. This paper reviews the current status of these activities, specifically the design and performance of the helium gas target cooling system.

07 ISOTOPE AND RADIATION SOURCES↗

Study on Shielding Requirements for Radioactive Waste Transportation in a Mo-99 Production Plant - 13382

Brazil is currently planning to produce {sup 99}Mo from fission of low enriched uranium (LEU) targets. The planned end of irradiation activity of {sup 99}Mo is about 185 TBq (5 kCi) per week to meet the present domestic demand of {sup 99m}Tc generators. The radioactive wastes from the production plant will be transferred to a waste treatment facility at the same site. The total activity of the actinides, fission and activation products present in the wastes can be predicted based on the yields of fission and activation data for the irradiation conditions, such as composition and mass of uranium targets, irradiation time, neutron flux, production schedule, etc., which were in principle already established by the project management. The transportation of the wastes from the production plant to the treatment facility will be done by means of special shielded packages. An assessment of the shielding required for the packages has been done and the results are presented here, aiming at contributing to the design of the waste management facility for the {sup 99}Mo production plant. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Evaluation of Novel Tin Materials for the Removal of Technetium from Groundwater

Technetium-99 ( 99 Tc) is present at several U.S. Department of Energy (DOE) facilities, including the Hanford, Oak Ridge, Paducah, Portsmouth, and Savannah River sites. Due to its mobility, persistence, and toxicity in the environment, developing means to immobilize and/or remove technetium from the environment is currently a top priority for DOE. However, there are currently very few approaches that effectively manage the risks of technetium to human health and the environment. The objective of this study is to evaluate novel synthetic materials that could enable direct removal of technetium from groundwater. The following report •assesses the viability of existing methodologies for synthesis of tin (II) apatite for in situ formation and remediation of 99 Tc within the subsurface environment •discusses the development of alternative methodologies for production of tin (II) apatite •evaluates nanoporous tin phosphate materials for removal of technetium from groundwater.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

ANSTO Synroc Wasteforms for the Immobilization of a Range of Mo-99 Production Wastes - 19326

Tc-99m, the most commonly used radioisotope in nuclear medicine, is produced from the decay of Mo-99 which is predominantly produced in reactors by irradiation of enriched uranium targets. This results in three waste streams, one of which is a uranium filter cake containing fission products. This filter cake could itself be treated as a waste stream, or processed to retrieve the enriched uranium for further use, thus producing additional waste streams during reprocessing. All wastes are considered intermediate-level wastes and require an appropriate wasteform design for termination. ANSTO is collaborating with NECSA (Nuclear Energy Corporation of South Africa) to develop wasteform and processing solutions for the treatment of intermediate level wastes (ILW) from the production of Mo-99. The work is supported by the U.S. Department of Energy's (DOE) National Nuclear Security Administration (NNSA) with technical support from Argonne National Laboratory (ANL). Here we highlight some of the outcomes of this collaboration which has seen the investigation and performance testing of a range of waste treatment technologies. Among the treatment options for the various waste streams, including the enriched uranium filter cake itself, were glass, glass-ceramic, ceramic and geopolymer wasteforms. All wasteforms were characterized (e.g., by X-ray diffraction (XRD) and scanning electron microscopy (SEM)) and their performance characteristics, for example compressive strength and aqueous durability, determined by standard methods and compared. Consideration has been given to the most suitable, practical and cost effective solutions for the immobilization of these Mo-99 wastes. This paper will discuss the down-selection of wasteforms and process technologies for Mo-99 wastes and compare these with regards to wasteform performance and process technology readiness. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Texas A&M Radioisotope Production and Radiochemistry Program

The main motivation of the project at Texas A&M University was to carry out the production of critically needed radioisotopes used in medicine for diagnostic and therapy, and to establish an academic program in radionuclide production and separation methods. After a lengthy battle with the Texas A&M University Radiation Safety Office, the Texas Department of State Health Services granted us a license for the production of radionuclides in July 2015, allowing us to work in earnest in our project objectives. Experiments began immediately after licensing, and we started the assembly and testing of our target systems. There were four analytical/theoretical projects and two experimental target systems. These were for At-211 production and for Zn- 62/Cu-62 production. The theoretical projects were related to the production of Mo-99/Tc-99m using (a) a subcritical aqueous target system and (b) production of Tc-99m from accelerator-generated Mo-99 utilizing a photon-neutron interaction with enriched Mo-100 targets. The two experimental projects were the development of targetry systems and production of At-211 and Zn-62/Cu-62 generator. The targetry system for At-211 has been tested and production of At-211 is chronic depending of availability of beam time at the cyclotron. The installation and testing of the targetry system for the production of Zn-62/Cu-62 has not been finalized. A description of the systems is described. The academic program in radionuclide production and separation methods was initiated in the fall of 2011; due to the lack of a radiochemistry laboratory, it was suspended. We expect to re-start the academic program at the Texas A&M Institute for Preclinical Studies under the Molecular Imaging Program.

43 PARTICLE ACCELERATORS↗

Development of a Type A Shielded Liquid Package - 18215

In the U.S., very few cost-effective certified packages exist for shipping radioactive Type A quantities of liquids. Two activities where the lack of such a package was identified as creating a serious need were: 1) shipping Mo-99, both across the United States for medical purposes; as well as, 2) shipping high Cs-137 concentration High Level Waste (HLW) supernatant from the Savannah River Site and/or the Hanford Site potentially on public roads and highways for detailed laboratory characterization. For shipping Mo-99, the need for a package was amplified by its ubiquitous use in the radio-medical industry. Although Tc-99m is used in approximately 80% of the medical radioisotope procedures, it is almost universally shipped as Mo-99, with the Mo-99 then converted to Tc-99m near the point of use (e.g. at the doctor's office or hospital). This transport of Mo-66 with the subsequent conversion to Tc- 99m is because Mo-99 has a 66-hour half-life, while Tc-99m has only a 6-hour half-life. With the noted need for packaging to support Mo-99 shipments, a new cost-effective drum based packaging which could transport Type A quantities of liquids (including those with a high specific activity/gamma energy) for use on public highways and commercial aircraft was developed. Starting with the well analyzed design and analysis of a physically similar 'inner 30 gallons, outer 50 gallons' packaging (i.e. DOE 9979 package) certified for Type AF quantities of solids, this paper discusses the additional design, regulatory requirements, and tests associated in certifying the Heavily-shielded Type A Liquid-package (HVTAL). Based on analyses, the authorized content of the package was then expanded for additional Type A non-fissile liquids, including HLW supernatant. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Medical Isotope Production Analyses In KIPT Neutron Source Facility

Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is driven by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99 Mo is the parent isotope of 99m Tc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Low Count and Background Radionuclides Analysis - 20488

The US Department of Energy (DOE) is often faced with the need to evaluate radionuclides at low concentrations. When site sample data are likely to be close to threshold activity concentrations of interest, then the means by which the radiochemical analysis is performed and reported is critical. This situation can occur when differentiating from zero (presence/absence) for radionuclides that do not occur naturally, close comparison with environmental background for naturally occurring radionuclides, close comparison with a risk- or dose-based threshold concentrations of interest, or even comparisons across studies. There are several analytical issues that are of concern, but the two that appear to cause incorrect decisions to be made most often involve establishing detection limits and subtracting ambient background conditions in the laboratory. These issues are not critical when radionuclide activity concentrations are large relative to thresholds of concern, but they seem to be poorly understood when it matters. When the comparisons are important and are likely to be close to a threshold of interest, then the general contract with the analytical laboratories needs to be changed so that the right or appropriate data are obtained. The concern is that important decisions are made incorrectly more often as greater scrutiny is placed on DoE's radionuclide cleanup or monitoring decisions by the public and other stakeholders. Examples are presented of problems that have been observed for different projects, both within and outside the realm of DOE and NRC remediation and radioactive waste disposal problems, and solutions are offered that should lead to better data from which important decisions need to be made. The first example is from Los Alamos National Laboratory (LANL) and involves radionuclide concentrations in soil and rock beneath LANL's Material Disposal Area (MDA) G. An initial review of the data led to a conclusion that americium and plutonium are a long way present beneath MDA G. A more thorough review of the data that accounted properly for ambient background and the detection limits that had been established led to the opposite conclusion. Another example is from the Nevada National Security Site where tritium results from one of the wells were unexpectedly high. Proper understanding and analysis of ambient background led to the conclusion that the increased concentrations were not so obvious, and that a different contract with the analytical laboratory was needed to provide more appropriate data to support a better determination. Other examples are used from regulatory review of projects in Nevada, where background levels and secular equilibrium for naturally occurring radionuclides are not established correctly because of analytical issues. The same basic issues have also been found to create difficulties analyzing historical data from the West Valley Demonstration Project. There is evidence in the data that the apparent lack of secular equilibrium where it is expected to exist is related to ambient background subtraction or other analytical issues. A final example is presented for analysis of Tc-99 in samples of depleted uranium. In this case, two different studies that were performed only three months apart provide quite different results. The US Environmental Protection Agency (EPA) established the data quality objectives (DQO) process in the mid-1980's to establish decision performance criteria for data collection. EPA guidance (EPA G-4, for example) clearly distinguishes between DQOs and measurement performance objectives (MQOs) that should be addressed for laboratory analysis of samples. The language of DQOs and MQOs has become confused over time it seems, and the subsequent effects seem to include a lack of attention to decision performance and a routine approach to measurement quality. In order to better address radionuclide sample analysis when the concentrations are close to thresholds of concern, which might be zero for some radionuclides, background for others, and risk-based thresholds for yet others, it is important that routine laboratory analysis methods are adjusted, and that the project team and the laboratory work closely together to ensure that the data meets the MQO requirements of laboratory analysis and reporting of results, and that the MQOs effectively support project-specific DQOs. This basic approach will be applied in Los Alamos in the coming year to the collection of moisture data from underneath MDA T that will be analyzed for americium and neptunium isotopes. Proper understanding of the radiochemistry methods and reporting, and of appropriate statistical methods is critical to the success of such projects, ensuring that the right decisions are made. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

Experiment on the treatment of waste extraction solvent from the molybdenum-99 process

In the Mo-99 (Molybdenum-99) isotope extraction test process for radiopharmaceutical applications, organic solvent is used to extract Mo-99 from an irradiated UO{sub 2} dissolution. The extraction solvent was stored when the test work was stopped. A total of about 120 liters of waste solvent was stored at INER (Institute of Nuclear Energy Research, Taiwan). The extraction solvent consisted of 5% di-(2-ethylhexyl)-phosphoric acid (D2EHPA) and kerosene. The radionuclides found in the waste solvent include Cs-137, Am-241, Tc-99, and Sr-90, which give off gross alpha and beta radioactivity of 1898 and 471 Bq/ml, respectively. This study aims to remove radionuclides from the waste solvent using sodium carbonate and sodium hydroxide solutions in different concentrations. After mixing the waste solvent with the alkaline solution followed by settling, a third phase other than organic and aqueous phase appeared which is expected due to the saponification reaction. The experimental results showed that increasing the number of washing and the alkaline solution concentration could enhance the radionuclides removal rate. An optimal removal method was proposed using 2 M Na{sub 2}CO{sub 3} solution twice followed by 1 M NaOH solution one time for the third phase generated early in the mixing stages. The remaining gross alpha and beta radioactivity of the treated organic solvent was 2 and 3 Bq/ml, respectively. The treated solvent could be stabilized by ashing at 500 deg. C and then immobilized. The alkaline solution would be neutralized by hydrochloric or nitric acid and then treated using a variety of adsorbents or bone char via adsorption to remove nuclides to meet the wastewater discharge limitation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of the Human Biohazard of Long-Lived Fission Products and Actinides for BREST-OD-300 Spent Fuel

The aim of this work is to perform a comparative analysis of the biohazard for the population of long-lived fission products and actinides from BREST-OD-300 reactor spent fuel taking account of the time required for different radionuclides to reach Earth’s surface by migrating from a deep disposal site. It was found that the biohazard of fission products is significantly greater than that of actinides for a long time (20000 years according to the accepted scenario of radionuclide migration). The radiological impact on the population is predicted after {sup 99}Tc – radionuclide with high migration capability – reaches the soil surface. {sup 137}Cs and {sup 90}Sr will not appreciably affect the dose load because of their relatively short half-life. In order to maintain radio-ecological balance between long-lived wastes and native uranium during the entire possible time period, it is sufficient to extract from the BREST-OD-300 spent fuel 90% of the technetium and 75% of the americium with 99.9% of the fissile materials extracted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Liquid Secondary Waste Grout Formulation and Waste Form Qualification

This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99 Tc desorption K d (distribution coefficient) values for 99 Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99 Tc(IV)-bearing solid phases for 99 Tc transport in reducing conditions to support the IDF PA.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

129 I, 99 Tc, and U Distribution Coefficients of Subsurface Sediments Collected from the Proposed Site of the Environmental Management Disposal Facility

Performance Assessment calculations were completed in 2020 to evaluate the Environmental Management Disposal Facility (EMDF), a proposed new low-level radioactive waste (LLW) disposal facility on the U.S. Department of Energy’s Oak Ridge Reservation (ORR). Among the large number of input parameters needed for such calculations, are distribution coefficients (K d values; radionuclide concentration solid: liquid ratio) that provide a measure of the tendency of radionuclides to bind to sediments. The objective of this study was to measure K d values of three radionuclides that may pose a disproportionately large amount of risk, U, iodine-129 ( 129 I) and technetium-99 ( 99 Tc). The average 129 I K d value for the 14 geological materials recovered from the proposed EMDF site was 37.8 mL/g and ranged from 0.45 to 140.9 mL/g. These values were consistent, but somewhat larger than previous measurements made with ORR sediments and were about an order of magnitude greater than those used in previous EMDF PA calculations. The median 99 Tc K d value was 365.7 mL/g, much greater than previously reported using ORR geological materials. Five of the 14 tested geological materials sorbed large quantities of 99 Tc, suggesting that the weakly sorbing 99 Tc(VII) species had been reduced to the sparingly soluble 99 Tc(IV) species. The five strongly sorbing sediments had apparent 99 Tc solubility values of approximately <10 -8 mol/L. The median U K d value was 5,726 mL/g. All of the tested geological materials had large K d values, ranging from 625 to >10,208 mL/g. Among the sediment samples that exhibited strong U binding, the apparent solubility value was approximately <10 -9 mol/L. Based on sediment properties and general ORR geological considerations, it was proposed that much of the 129 I and 99 Tc retention could be attributed to the site materials exhibiting low pH (average pH = 4.94), and/or the elevated levels of iron oxides, manganese oxides, and natural organic matter. Similarly, the extremely high U binding measured in these sediments may also be attributed to the low conditions of carbonates, which can complex and therefore solubilize uranyl in these tests due to the low pH, and also the relatively high concentrations of iron and organic coatings on these samples. An implication of this study is that the areas of the EMDF subsurface environment may have natural properties for attenuating 129 I, 99 Tc, and U movement, and potentially other radionuclides, thereby possibly reducing risk posed by burial of LLW at this site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Effect of Ettringite on Cementitious Waste Form Development for Low- Activity Waste - 17170

The current Cast Stone formulation, with 8% ordinary Portland cement (OPC), 45% fly ash (FA), and 47% blast furnace slag (BFS), proposed for cementitious waste forms (CWFs) at the Hanford site has been shown to be inadequate for secondary liquid wastes sent to the Effluent Treatment Facility (ETF). This is due to the high sulfate concentrations and non-alkaline conditions present in the ETF waste stream that lead to the late formation of ettringite, [Ca{sub 6}Al{sub 2}(SO{sub 4}){sub 3}(OH){sub 12}.26(H{sub 2}O)], which can consequently cause undesirable swelling and cracks in the CWF that facilitate {sup 99}Tc release. To overcome this challenge, hydrated lime (HL), Ca(OH){sub 2}, can be used instead of FA to provide the Ca necessary to sequester sulfate early in the cast stone curing process through early formation of ettringite. Early ettringite formation minimizes the risk of localized disruptive action or harmful stress that lead to undesirable swelling and cracking of the CWF. Furthermore, early formation reduces the permeability of the CWF, further reducing the risk of {sup 99}Tc release and increasing structural stability. Two liquid waste simulants, 1) WTP off-gas condensate (WTP) and 2) Effluent Management Facility (EMF) process condensate combined with low activity waste (LAW) caustic scrubber (labeled EMF/CS), were prepared and tested for ettringite formation effects on HL-containing CWF development and {sup 99}Tc leachability. After curing for 28 days, X-ray diffraction analysis revealed that ettringite was formed in the early stages of curing for the CWFs prepared using both simulants. However, because of the higher sulfate concentration in the EMF/CS simulant and limited Ca and Al sources, some sulfate did not form ettringite during the early stages of curing, resulting in significant monolith swelling and cracking during subsequent leaching tests. Despite observable cracking in monoliths prepared with the EMF/CS simulant, low {sup 99}Tc release from HL-based waste forms was measured in later leaching periods tested by the EPA Method 1315 in both simulants. Ettringite-{sup 99}Tc co-precipitation batch experiments support a possible chemical sequestration mechanism that is hypothesized to contribute to the observed low {sup 99}Tc leach rates, which are likely the result of an additional chemical reaction that sequesters {sup 99}Tc rather than simply physical factors such as reduced porosity and permeability (or tortuosity). Overall these findings quantitatively support the use of HL to overcome the deleterious effects of sulfate-rich conditions on CWFs through early formation of ettringite and provide evidence for {sup 99}Tc release rates that fall within low risk levels reported by the United States Department of Energy (US DOE). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

ELECTRON-CAPTURE SUPERNOVAE AS SOURCES OF {sup 60}Fe

We investigate the nucleosynthesis of the radionuclide {sup 60}Fe in electron-capture supernovae (ECSNe). The nucleosynthetic results are based on a self-consistent, two-dimensional simulation of an ECSN as well as models in which the densities are systematically increased by some factors (low-entropy models). {sup 60}Fe is found to be appreciably made in neutron-rich ejecta during the nuclear quasi-equilibrium phase with greater amounts being produced in the lower-entropy models. Our results, combining them with the yields of core-collapse supernovae in the literature, suggest that ECSNe account for at least 4%-30% of live {sup 60}Fe in the Milky Way. ECSNe co-produce neutron-rich isotopes, {sup 48}Ca, {sup 50}Ti, {sup 54}Cr, some light trans-iron elements, and possibly weak r-process elements including some radionuclides such as {sup 93}Zr, {sup 99}Tc, and {sup 107}Pd, whose association with {sup 60}Fe might have been imprinted in primitive meteorites or in the deep ocean crust on the Earth.

79 ASTRONOMY AND ASTROPHYSICS↗