OBJECTIVES/GOALS: A quarter of Puerto Rican pediatric population are overweight or obese. Pediatric obesity is established as a global public health concern that will develop liver complication as a Non-alcoholic fatty liver disease (NAFLD). This study aims to determine the association between body weight, liver size and texture in Puerto Rican pediatric population. METHODS/STUDY POPULATION: Twenty-nine (n=29) pediatric participants (20 unhealthy weight and 9 healthy weight) between 7 to 19 years underwent panoramic ultrasound imaging of the RLL. Craniocaudal RLL measurement and liver texture were evaluated. Body mass index and waist circumference were also compared. Shapiro-Wilk test and students t-tests were attained with significance at p < 0.05. RESULTS/ANTICIPATED RESULTS: Statistical differences were detected between healthy weight and unhealthy weight pediatric patients: (1) RLL craniocaudal diameter (p=0.006); (2) waist circumference (p < 0.001); (3) BMI (p < 0.001). Unhealthy weight (overweight and obese) pediatric patients presented a greater number of livers with fat infiltration (13) and hepatomegaly (15). DISCUSSION/SIGNIFICANCE: This preliminary results showed that liver size was larger and fatty infiltration are relatively frequent among overweight and obese pediatric patients. Prevention of unhealthy weight and the diagnosis of changes in liver texture and size among pediatric patients is important to prevent progressing of liver disease and avoid irreversible damage.

The release of radionuclides measured during washing cycles of spent nuclear fuel samples in a series of experiments using different solutions are analyzed with respect to the fission products Cs, Sr, and Tc and the actinides U, Pu, and Am. Based on the concentrations of the dissolved radionuclides, their release rates are evaluated in terms of fraction of inventory in the aquatic phase per day. The application of this information on the fast/instant radionuclide release fraction (IRF) is discussed and following issues are addressed: Duration of the wash steps, solution chemistry, and radionuclide sorption onto surface of the experimental vessels. Data for the IRF are given and the correlation between the mobilization of the various elements is analyzed.

In accordance with the Belgian “supercontainer design”, spent nuclear fuel (SNF) will be encapsulated in carbon steel canisters, surrounded by a concrete overpack for disposal in poorly-indurated clay. After re-saturation of the barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Corrosion studies of SNF in ECW-type solution (Evolved Cement Water) and YCWCa-type solution (Young Cement Water with Ca) were performed under externally applied H2 overpressures over 426 days. Directly after H2 application, Tc concentrations decreased from >10-8 M to concentrations below detection limit. Based on the fractional release of selected fission products, low matrix dissolution rates of ~10-8/day were found in both experiments. U concentrations decreased finally to 1.5•10-9 M (YCWCa) and to 2.1•10-10 M (ECW), respectively. Am, Np and Pu concentrations were found throughout the experiments below their detection limits indicating an effective retention process.

This study is focused on the alteration behavior of spent nuclear fuel when exposed to highly alkaline groundwater. Contact of highly alkaline solution with the waste product is considered in the Belgian concept for disposal in the Boom Clay formation. According to the “supercontainer design” the fuel will be encapsulated in carbon steel canisters, surrounded by a concrete over-pack. After saturation of the engineered barriers by porewater, interactions with the concrete will result in solutions rich in NaOH, KOH and Ca(OH)2. Using this type of solution at pH 12.5, spent nuclear fuel corrosion experiments were conducted over 378 days. Under anoxic conditions, parallel experiments were performed (a) in the absence of Fe phases and (b) in the presence of solid Fe phases representing container (corrosion) products. Both types of experiments resulted in relatively low matrix dissolution rates, around 10-7 per day, according to the fractional release of Sr. Solution concentrations of actinides are close to or below the detection limit, indicating an effective retention of these radioelements in the system. The observed precipitation of a Ca rich phase onto the surfaces of the corroded fuel samples may be related to the inhibited re-lease of actinides, Sr and other matrix bound radioelements.

Recent studies have shown that in the presence of H2 overpressure, which forms due to the corrosion of the Fe based container, the dissolution rate of the spent fuel matrix is slowed down by a factor of about 10, associated with a distinct decrease of concentrations of important ra-dionuclides. However, in a natural salt environment as well as in geological formations with chloride rich groundwater the presence of radiation chemically active impurities such as bro-mide must be taken in consideration. Bromide is known to react with β/γ radiolysis products, thus counteracting the protective H2 effect. In the present experiments using high burnup spent fuel it is observed that during 212 days the matrix dissolution rate was enhanced by a factor of about10 in the presence of up to 10-3 M bromide and 3.2 bar H2 overpressure. However, concen-trations of matrix bound actinides were found at the same level or below as found under identical conditions, but in the absence of bromide. In the long-term it is expected that the effect of bro-mide becomes less important, because the decrease of β/γ-activity results in a decrease of oxida-tive radicals, which react with bromide, while α activity will dominate the radiation field.

The long-term immobilization for individual radioelements released from the waste form “spent fuel” in solid phases upon groundwater contact depends strongly on the (geo)chemical constraints prevailing in the repository. Related experimental studies comprise effects induced by the presence of Fe based container material, and near field materials other than Fe for a rock salt environment. The effect of the presence of an argillaceous host rock containing organic matter and pyrite on fuel alteration was studied in addition. The results have shown that oxidative radio-lysis products were found to be consumed at a significant extent by the metallic Fe and by the argillaceous host rock. Under these conditions a decrease at a factor of ca.100 for both the matrix dissolution rates and the solution concentrations of U and Pu was found. There is mutual support between the matrix dissolution rates, the solution concentrations and the amounts of oxygen encountered during the experiments under various conditions controlled by the presence of near field materials under study.

The corrosion of the Fe-based spent fuel containing canister produces large amounts of hydrogen, which dominate the composition of the gas phase. To quantify to what extent the hydrogen over- pressure may counteract radiolysis enhanced matrix dissolution, related experimental work has been performed. High burnup spent fuel was corroded in 5 M NaCl solution in different tests: (1) Fe ab-sence (<0.16 bar H2 overpressure), (2) Fe presence (2.8 bar H2 overpressure), (3) Fe absence, external applied H2 overpressure of 3.2 bar. In the absence of Fe, after the application of H2 overpressure the UO2 matrix dissolution rate decreased by a factor of about 10. The concentrations of U, Np, Tc in so- lution were found to be decreasing at least two orders, and ranging within the same level as in the presence of Fe powder. However, Pu and Am concentrations were less affected, due to the absence of Fe powder and the associated high sorption capacity for these radioelements.

During long-term interim storage of spent fuel, pre-oxidation of the UO2-matrix may not be ruled out completely. This can happen if air could find access to the fuel in the case of cladding failure. The aim of this work is to study the impact of pre-oxidation of the fuel surface on the UO2 matrix dissolution rate and the associated mobilization or retention of radionuclides in highly concentrated salt solutions. The tests were performed with samples that suffered pre-oxidation during up to seven years. The dissolution rate of a fuel sample contacted by small quantities of air-oxygen was found to be roughly a factor of 10 higher in comparison to non oxidized samples, but concentrations of radionuclides, especially Pu and U were hardly affected. The majority of dissolved radionuclides, especially Pu, U appear to have been reimmobilized on the fuel sample itself.

With respect to the assessment of the long-term behavior of the waste form spent fuel it is of high importance to study the fuel alteration in contact with groundwater and near field materials. The aim of this work is to evaluate the impact of candidate backfill materials hydroxylapatite and magnetite on the overall corrosion behavior of this waste form in salt brine; both materials are used in corrosion tests together with spent fuel. The instant releases and the matrix dissolution rates appear to be similar in presence and in absence of any backfill material under study. However, Am,Np,Pu,U and Sr are retained at different ratios on the hydroxylapatite, on the magnetite and on the fuel sample, indicating possibly the formation of different radionuclide containing new solid phases.

With respect to the assessment of the long-term behavior of the waste form spent fuel it is of high importance to study in particular the UO2 matrix dissolution behavior and the associated release/retention of radionuclides in contact with groundwater. During long term fuel storage, fuel oxidation may not be avoided. Main issue of this work is to identify the impact on the corrosion of partly oxidized fuel of environmental conditions such as (1) the nature of solution contacting the matrix, the (2) presence/absence of CO2, (3) fixed pH values within a range between pH 7- pH 11, and (4) the presence/absence of corroding container material (Fe-powder). Dissolution tests with powdered oxidized spent fuel in various granite waters, and NaCl-brine resulted in matrix dissolution rates in the same order of magnitude for all investigated media (ca.5×10−4/d). The presence of CO2 and fixed pH values (pH 5 – 11) was without a distinct effect. The independence of the dissolution rate of the oxidized fuel matrix upon the nature of solution, pCO2, fixed pH values (5-11) can probably be explained by a masking effect of radiolysis. In presence of Fe powder the matrix dissolution rate was found to be slowed down by a factor of ca. 20, associated with strong retention effects of radionuclides.

In Germany certain quantities of spent MOX fuel will be disposed in a rock salt repository. In order to assess the safety of the repository the corrosion behavior of spent MOX fuel in salt solution must be understood. Since material properties of spent MOX fuel are different compared to these of spent LWR UO2 fuel, it is not possible to deduce from corrosion behavior of the latter to that one of spent MOX fuel. It is expected that the most significant differences will occur in the dissolution kinetics, whereas similar solid alteration products are expected from the similar overall composition of the fuels. Various experiments were performed focused on the effect of fuel type on the dissolution kinetics. The results of this work are summarized and evaluated

Dissolution of spent fuel has been studied in saline, anaerobe, carbonate free solutions. Processes controlling spent fuel dissolution and associated radionuclide release are radiolytically controlled oxidative dissolution, sorption on container, solubility and coprecipitation. Upper limits for oxidative dissolution rates are given by the production rates of oxidative radiolysis products. This limitation leads to a strong decrease in surface area normalized reaction rates with increasing surface to volume ratio (S/V) and imposes geometric constraints on prediction of spent fuel behavior in a repository. Solution concentrations of Am during spent fuel corrosion were about 5 orders of magnitude lower than the solubility of Am(OH)3(s) and are likely controlled by coprecipitation. Pu concentrations may be controlled by Pu(VI) or Pu(IV) (hydr)oxides.

The objective of this investigation is to describe the extent to which Np, Pu, Am and Tc are mobilized from vitrified high-level radioactive waste into the near field of an HLW repository in a salt formation, when a hot and concentrated salt solution comes into contact with the glass. Waste form corrosion studies are conducted with a salt solution representing the composition of a fluid phase encountered in drill holes in the Gorleben salt dome. Test temperatures are determined by the designed maximum surface temperature of 200°C for the vitrified waste in the Gorleben salt. The following results were obtained: 1. pH changes of the radio-active leachate are the same as in inactive leachates. 2. The time and temperature dependence of the reaction for the radioactive glass are in excellent agreement with that of the inactive glass. 3. Np, Pu, Am, and Tc have not been reimmobilized in secondary minerals. Hence, mobilization of these radionuclides is governed by the kinetics of glass dissolution. Pu oxidation states were analyzed and related to Pu concentrations.

High-burnup (<50 MWd/kgU) spent fuel samples of various sizes were exposed to NaCl solutions under static, anaerobic and reducing conditions. The accumulated corrosion time was about 200 days. Gas phase and leach solutions were analyzed. By dissolving mm sized fragments in large volumes of solution, saturation effects were avoided and upper limits for intrinsic dissolution rates of about 5-20 mg/(m2d) were measured. Surface area normalized reaction rates were significantly lower when using fine grained fuel powder (estimated sample surface area to solution volume ratio S/V ca. 3000 m-1), indicating saturation effects. The maximum concentrations of Pu and Am in the tests are close to reported solubility limited concentrations in pure 5m NaCl solutions in the absence of radiolysis effects. The presence of iron effectively reduces the solution concentration of all measured radionuclides (except Cs).

Aluminum silicate based ceramics were studied as matrix materials for the immobilizationof real TRU—wastes, as well as simulated high-level and TRU—waste mixtures. As observed by XRD and electron microscopy, isolation of radionuclides is achieved either by encapsulation of radiophases by aluminumsilicate matrix phases and by incorporation into host phases in the matrix. TRU—leach rates were found to be on the order of 5*10–3 and 1*10–2 gm–2 d–1 at room temperature after one year in MgCl2 —rich salt solution. By doping the ceramic matrix with 20 weight percent PuO2 an accumulated α-dose of 8.33*109 Gy (1*1019 α-decays*g–1) was reached after 3 years. The microstructure and the lattice constants ofthe matrix phases were not affected by a-irradiation.