Recovery of uranium from seawater.

A computer program entitled URPE (Uranium Recovery Performance and Economics) has been developed to simulate the engineering performance and provide an economic analysis O of a plant recovering uranium from seawater. The conceptual system design used as the focal point for the more general AM analysis consists ofa floating oil-rig type platform, Asinlge-point moored in an open ocean current, using either high volume, low head, propeller pumps or the velocity head 4M of the ambient ocean current to force seawater through a mass transfer medium (hydrous titanium oxide (HTO) coated onto particle beds or stacked tubes), as in most process designs previously suggested for this service. Uranium is recovered Sfrom the seawater by an adsorption process, and later eluted . from the adsorber by an ammonium carbonate solution. A multi-product co-generating plant on board the platform burns coal to raise steam for electricity generation, desalination, and process heat requirements. Scrubbed stack gas from the plant is processed to recover carbon dioxide for chemical make-up needs.The equilibrium isotherm and the diffusion constant for the uranyl-HTO system, which are needed for bed performance calculations, have been calculated based on the rather sparse data reported in the literature. In addition, a technique for calculating the rate constant of a fixed bed adsorbing system has been developed for use with Thomas' solution for predicting fixed bed performance.The URPE program has been benchmarked against the results of previous studies by ORNL and Exxon, and found to make comparable performance and economic estimates when applied under the same set of ground rules. The URPE code was then used in an extensive series of parametric and sensitivity studies to identify optimum bed operating conditions and important areas for future research and development. The program showed that thin beds of small, thinly-coated particles were the preferred bed configuration, and that actively pumped systems out-perform current driven units.Based on the URPE analysis, the minimum expected costs nof uranium recovered from seawater would be no lower than ~316 (1979$)/lb U308 for state-of-the-art adsorber material (capacity equal to 210 mg U/kg Ti), but might be reduced to the level of breakeven attractiveness of ~150 (1979$)/lb U30 8 if at least a four-fold increase in adsorption capacity could be achieved. Specific research and development objectives other than increasing particle capacity are also indentified. Prospects are considered to be sufficiently good to warrant recommending further work

Systems studies on the extraction of uranium from seawater

This report summarizes the work done at MIT during FY 1981 on the overall system design of a uranium-from-seawater facility. It consists of a sequence of seven major chapters, each of which was originally prepared as a stand-alone internal progress report. These chapters trace the historical progression of the MIT effort, from an early concern with scoping calculations to define the practical boundaries of a design envelope, as constrained by elementary economic and energy balance considerations, through a parallel evaluation of actively-pumped and passive current-driven concepts, and thence to quantification of the features of a second generation system based on a shipboard-mounted, actively-pumped concept designed around the use of thin beds of powdered ion exchange resin supported by cloth fiber cylinders (similar to the baghouse flyash filters used on power station offgas).An assessment of the apparently inherent limitations of even thin settled-bed sorber media then led to selection of an expanded bed (in the form of an ion exchange "wool"), which would permit an order of magnitude increase in flow loading, as a desirable advance. Thus the final two chapters evaluate ways in which this approach could be implemented, and the resulting performance levels which could be attained. Overall, U 308 production costs under 200 $/lb appear to be within reach if a high capacity (several thousand ppm U) ion exchange wool can be developed

Interfacial effects in fast reactors

"May 1979."Also written as a Ph. D. thesis by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1979Includes bibliographical references (pages 191-193)The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed to measure U-238 capture rates near the blanket-reflector interface in the MIT Blanket Test Facility. Prior MIT experiments on a thorium-uranium interface in a blanket assembly were also reanalyzed. Extremely localized fertile capture rate increases of on the order of 50% were measured immediately at the interfaces relative to extrapolation of asymptotic interior traverses, and relative to state-of-the-art (LIB-IV, SPHINX, ANISN/2DB) calculations which employ infinite-medium self-shileding throughout a given zone. A method was developed to compute a spatially varying background scattering cross section per absorber nucleus, a , which takes into account both homogeneous and heterogeneous effec~s on the interface flux transient. This permitted use of the standard self-shielding factor method (Bondarenko f-factors) to generate modified cross sections for thin layers near the interfaces. Calculations of the MIT experiments using this approach yielded good agreement with the measured data.U.S. Department of Energy contract EY-76-S-02-225

Evaluation of high performance LMFBR blanket configurations

Substantially the same as a Ph. D. thesis by G.J. Brown in the Dept. of Nuclear Engineering, MIT, 1974Includes bibliographical references (pages 249-254)AEC Contract AT(11-1)-225

Heterogeneous effects in fast breeder reactors

"January, 1973."Also issued as a Ph. D. thesis written by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1973Includes bibliographical references (pages 259-266)Heterogeneous effects in fast breeder reactors are examined through development of simple but accurate models for the calculation of a posteriori corrections to a volume-averaged homogeneous representation. Three distinct heterogeneous effects are considered: spatial coarse-group flux distribution within the unit cell, anisotropic diffusion, and resonance self-shielding. An escape/transmission probability theory is developed which yields region-averaged fluxes, used to flux-weight cross sections. Fluxes calculated by the model are in substantial agreement with S 8 discrete ordinate calculations. The method of Benoist, as applied to tight lattices, is adopted to generate anisotropic diffusion coefficients in pin geometry. The resulting perturbations from a volume-averaged homogeneous representation are interpreted in terms of reactivities calculated from first order perturbation theory and an equivalent "total differential of k" method.Resonance self-shielding is treated by the f-factor approach, based on correlations developed to give the self-shielding factors as functions of one-group constants. Various reference designs are analyzed for heterogeneous effects. For a demonstration LMFBR design, the whole-core sodium void reactivity change is calculated to be 90e less positive for the heterogeneous core representation compared to a homogeneous core, due primarily to the effects of anisotropic diffusion. Parametric studies show at least a doubling of this negative reactivity contribution is attainable for judicious choices of enrichment, lattice pitch and lattice geometry (particularly the open hexagonal lattice). The fuel dispersal accident is analyzed and a positive reactivity contribution due to heterogeneity is identified. The results of intra-rod U-238 activation measurements in the Blanket Test Facility are analyzed and compared to calculations.This activation depression is found to be of the order of 10% (surfaceto- average) for a typical LMFBR blanket rod and is ascribed to the effect of heterogeneous resonance self-shielding of U-238. Heterogeneous effects on the breeding ratio are studied with the conclusions that accounting for resonance self-shielding reduces the total breeding ratio by over 10%, but heterogeneous effects are not important for breeding ratio calculations.U.S. Atomic Energy Commission contract AT (11-1) - 225

Some applications of Ge(Li) gamma-ray spectroscopy to fuel element assay

"MIT-3944 -5."Also issued as a Ph. D. thesis by the first author and supervised by the second and third author, MIT Dept. of Nuclear Engineering, 1970Includes bibliographical references (pages 195-198)It was the object of this work to study the gamma rays emitted by the products of the interaction of thermal neutrons with the nuclei of U-238, Th-232, U-235 and Pu-239 during and after irradiation and to explore some applications mainly to fuel element assay. An irradiation facility and a Ge(Li) detector cryostat were constructed for this purpose. A new method of assaying a fuel rod containing a mixture of plutonium and uranium oxide, based on the difference in the observed yield of the fission products 1-135 and Sr-92, has been developed. The energies and intensities of the thermal neutron capture gamma rays for U-238 and Th-232 were determined. Four new lines have been found in the energy region previously unexplored for U-238. For Th-232, 66 certain lines were found, compared to 7 lines in the literature. Many prompt gammas emitted 'by the highly excited fission products following the fission of U-235 and Pu-239 were resolved in the energy region above 1.4 MeV. For U-235 fissions, 57 lines were found, and for Pu-239, 51 certain lines were recorded. The use of prompt gammas for assaying fuel rods was investigated. An accuracy of about ± 7% was obtained for the analysis of U-238 content; ± 10% to ± 20% accuracy was obtained for U-235 analysis in the range of 1% to 2% enrichment; and ± 35% accuracy for the analysis of 0.25% Puenriched rods. It has been found that Ge(Li) detectors can be operated as fast neutron detectors and used to determine the relative neutron yield. With this method, the enrichment of uranium rods can be found with an accuracy of ± 1% to ± 2% in the range from 116 to 2% enrichment. Finally, some considerations were given to the use of prompt gamma rays for measuring the initial conversion ratio C and the neutron yield parameter [eta].U.S. Atomic Energy Commission contract AT(30-1)-394

An evaluation of tight-pitch PWR cores

Originally presented as the author's thesis, Ph.D. in the M.I.T. Dept. of Nuclear Engineering, 1979.The impact of tight pitch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system-U-235/U02: Pu/Th02: U-233/ThO2--and the conventional recycle-mode uranium system- U-235/U02: Pu/UO . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) o the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices)< F/M < 4.0 are limited by the scarcity of experiments with F/M > l.0,the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments. It was found that by increasing F/M to "3 the uranium ore usage for the uranium system can be decreased by as much as 60% compared to the same system with conventional recycle (at F/M 0.5). Equivalent savings for the thorium system of the type examined here are much smaller (10%) because of the poor performance of the intermediate Pu/ThO2 core--which is not substantially improved by increasing F/M. Although fuel cycle costs (calculated at the indifference value of bred fissile species) are rather insensitive to the characteristics of the tight pitch cores, system energy production costs do not favor the low discharge burnups which might other- wise allow even greater ore savings (80%). Temperature and void coefficients of reactivity for the tight pitch cores were calculated to be negative. Means for implementing tight lattice use were investigated, such as the use of stainless steel clad in place of zircaloy; and alternatives achieving the same objective were briefly examined, such as the use of D20/H20 mixtures as coolant. Major items identified requiring further work are system redesign to accommodate higher core pressure drop, and transient and accident thermal-hydraulics.DOE Contract no. EN-77-S O2-4570

Uranium utilization in PWRs.

This is the final summary progress report on a research program' carried out within the MIT Energy Laboratory/Nuclear Engineering Department under the US Department of Energy's program to increase the effectiveness of uranium utilization in light water reactors on the once-through fuel cycle.Two major themes, methodology and applications, characterize the research. A simple buit. accurate set of algorithms, designated as "the linear reactivity method" were developed to permit self-consistent evaluations of a broad spectrum of changes in core design and fuel management tactics.More than a dozen suggested improvements were then evaluated, focusing primarily on retrofitable modifications and pressurized water reactors. In common with the findings of many other investigators, high burnup and routine end-of-cycle coastdown were identified as preferred options.Division of Energy Technology, U.S. Dept. of Energy