[en] The caustic precipitation of plutonium (Pu) and uranium (U) from Pu and U containing waste solutions has been investigated to determine whether gadolinium (Gd) could be used as a neutron poison for precipitation with greater than a fissile mass containing both Pu and enriched U. Precipitation experiments were performed using both actual samples and simulant solutions with a range of 2.6-5.16 g/L U and 0-4.3 to 1 U to Pu. Analyses were performed on solutions at intermediate pH to determine the partitioning of elements for accident scenarios. When both Pu and U were present in the solution, precipitation began at pH 4.5 and by pH 7, 99 percent of Pu and U had precipitated. When complete neutralization was achieved at pH greater than 14 with 1.2 M excess OH-, greater than 99 percent of Pu, U, and Gd had precipitated. At pH greater than 14, the particles sizes were larger and the distribution was a single mode. The ratio of hydrogen to fissile atoms in the precipitate was determined after both settling and centrifuging and indicates that sufficient water was associated with the precipitates to provide the needed neutron moderation for Gd to prevent a criticality in solutions containing up to 4.3 to 1 U to Pu and up to 5.16 g/L U

[en] A general purpose dynamic optimization scheme suitable for high level waste (HLW) complex operations has been developed. The optimizer is interfaced with the SPEEDUPTM based dynamic simulator ProdMod for flow of information between the optimizer and simulator, while the optimization is performed in the stand-alone FORTRAN based optimizer. The linear constructs and the mapping algorithm of the ProdMod have been used in the optimization scheme for the interface. The optimization scheme has been successfully implemented in generating waste blending batch sequences for one of salt processing options at the Savannah River Site (SRS) HLW complex. Parametric studies demonstrate that the devised optimization scheme is a realistic approach for guiding the operation of the HLW complexes

[en] The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled, high temperature, fast neutron flux, loop-type test reactor. The facility was constructed to support development and testing of fuels, materials and equipment for the Liquid Metal Fast Breeder Reactor program. FFTF began operation in 1980 and over the next 10 years demonstrated its versatility to perform experiments and missions far beyond the original intent of its designers. The reactor had several distinctive features including its size, flux, core design, extensive instrumentation, and test features that enabled it to simultaneously carry out a significant array of missions while demonstrating its features that contributed to a high level of plant safety and availability. FFTF is currently being deactivated for final closure

[en] After closure of an underground nuclear waste repository, the decay of radionuclides will raise temperature in the repository, and the bentonite buffer will resaturate by water inflow from the surrounding host rock. The perturbations from these thermal and hydrological processes are expected to dissipate within hundreds to a few thousand years. Here, we investigate coupled thermal-hydro-chemical processes and their effects on the short-term performance of a potential nuclear waste repository located in a clay formation. Using a simplified geometric configuration and abstracted hydraulic parameters of the clayey formation, we examine geochemical processes, coupled with thermo-hydrologic phenomena, and potential changes in porosity near the waste container during the early thermal period. The developed models were used for evaluating the mineral alterations and potential changes in porosity of the buffer, which can affect the repository performance. The results indicate that mineral alteration and associated changes in porosity induced by early thermal and hydrological processes are relatively small and are expected to not significantly affect flow and transport properties. Chlorite precipitation was obtained in all simulation cases. A maximum of one percent volume fraction of chlorite could be formed, whose process may reduce swelling and sorption capacity of bentonite clay, affecting the performance of the repository. llitisation process was not obtained from the present simulations.

[en] Integrity and durability of solid radioactive wastes are related principally to rates at which waste constituents are leached into environmental water. This paper discusses a new and improved procedure which was developed for determining leachabilities of proposed radioactive waste forms. While no laboratory procedure can be expected to duplicate all possible environmental conditions, a single test is desirable to compare the leaching properties of solid waste forms produced in different laboratories

[en] We describe a joint inversion approach that combines geophysical and thermal-hydrological data for the estimation of (1) thermal-hydrological parameters (such as permeability, porosity, thermal conductivity, and parameters of the capillary pressure and relative permeability functions) that are necessary for predicting the flow of fluids and heat in fractured porous media, and (2) parameters of the petrophysical function that relates water saturation, porosity and temperature to the dielectric constant. The approach incorporates the coupled simulation of nonisothermal multiphase fluid flow and ground-penetrating radar (GPR) travel times within an optimization framework. We discuss application of the approach to a large-scale insitu heater test which was conducted at Yucca Mountain, Nevada, to better understand the coupled thermal, hydrological, mechanical, and chemical processes that may occur in the fractured rock mass around a geologic repository for high-level radioactive waste. We provide a description of the time-lapse geophysical data (i.e., cross-borehole ground-penetrating radar) and thermal-hydrological data (i.e., temperature and water content data) collected before and during the four-year heating phase of the test, and analyze the sensitivity of the most relevant thermal-hydrological and petrophysical parameters to the available data. To demonstrate feasibility of the approach, and as a first step toward comprehensive inversion of the heater test data, we apply the approach to estimate one parameter, the permeability of the rock matrix

[en] The reduction in turbulent, convective heat transfer parameters observed in some supercritical data and in experiments with common gases can be due to radial property variation, acceleration, buoyancy or combinations of these phenomena, depending on the conditions of the applications. To date criteria for the onsets of these effects have been developed for vertical circular tubes. This note presents extensions of these criteria to non-circular ducts with constant cross-sections as in the cooling channels of some advanced nuclear reactors

[en] A novel process to treat used heavy water moderator (D2O) contaminated with high concentrations of the neutron poison, gadolinium nitrate,Gd(NO3)3, is described. Gadolinium is removed by precipitation. The resultant precipitate (GdPO4. 6H2O) represents an extremely rare compound of considerable potential value. The resultant supernatant consisting of residual nitrate (NaNO3 or KNO3) is less toxic and easier to process than the original waste. Thus, the alkali metal waste handling can be done with considerably less environmental concern. This waste can potentially be treated by a combination of electrochemical and biological methods

[en] The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/L [U] and 1 M [H⁺] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 L of product solution, which was over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H⁺] in the range 0.8-1.2 M. A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of about 5000 ηCi/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO₂ present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U

[en] In this paper the applicability and usefulness of a complete and well-qualified plant transient code and model to support in-depth evaluation of anomalous plant transients are described. The qualified best-estimate RETRAN-02 model for the Cofrentes nuclear power plant (a boiling water reactor with an uprated power of 2952 MW) has been updated for RETRAN-03 using algebraic slip and one-dimensional kinetics. This model has been used in the analysis of recent abnormal plant transients at Cofrentes, including a partial control rod insertion at 92% power, a turbine trip at 67% power with reactor vessel overfill, and reactor instabilities during startup