Reprocessing Of Fuels Research Articles

Selected conversion of NOx by catalytic reduction with ammonia. An effective off‐gas treatment system is an environmental must in the reprocessing of nuclear fuels. An up‐to‐date progress report

The General Atomic Company off-gas treatment system is designed to process simulated radioactive or other noxious volatile and gaseous constituents in dissolver off-gas (DOG) streams indigenous to several nuclear fuel cycles, e.g., light-water reactors (LWRs). A series of NO/sub x/ converter tests was carried out with the GA engineering-scale system. The major reactions and optimum conditions were characterized for the selective catalytic reduction of pure NO, pure NO/sub 2/, and NO/sub x/ with NH/sub 3/ over synthetic hydrogen mordenite (Zeolon 900 H). Steady-state conversions in the range 99 to 99.9% were obtained for the specified conditions without observing any breakthrough of NH/sub 3/ in the effluent. The NO/sub x/ reduction capacity for a single-stage adiabatic converter is approximately 1% NO/sub x/ in the feed gas and is restricted by catalyst temperature limitations. 7 references, 9 figures. (JMT)

Leaching Properties and Chemical Composition of Calcines Produced at the Idaho Chemical Processing Plant

Aqueous wastes from the reprocessing of nuclear fuels at the Idaho Chemical Processing Plant are solidified in the Waste Calcination Facility. The calcined wastes are then stored in stainless steel bins contained in concrete vaults. It has been determined that the radionuclides of cesium and strontium leach readily on contact with aqueous solutions from the particulate fresh product calcine. Thus, the calcine may require further treatment for final disposal. Calcine that is stored up to 12 years has remained particulate and its chemical and other physical properties are similar to those of the original product calcine.

Radionuclides in Pronghorn Resulting from Nuclear Fuel Reprocessing and Worldwide Fallout

Radionuclides in Pronghorn Resulting from Nuclear Fuel Reprocessing and Worldwide Fallout

German nuclear reprocessing: Decision on site

German nuclear reprocessing: Decision on site

Ruthenium in Fuel Reprocessing by Fluoride Volatility Process, (II)

The following three reactions were studied by thermochromatography in relation to the reprocessing of oxide fuels: (a) fluorination of RuF3 by F2, (b) fluorination of RuO2 by F2, (C) fluorination of RuO2 mixed with an excess of an oxide UO2 or Y2O3. The reaction products were as follows. The RuF3-F2 reaction yields RuF5. Fluorination of RuO2 by F2 results in the production of a mixture of RuF4-RuF5-RuOF4. (RuF5 may be produced by fluorination of RuF4 resulting from decomposition of RuOF4 in the apparatus for thermo-chromatography). When dried at 150°C prior to fluorination the products of RuO2-F2 reaction in the presence of an excess UO2 on Y2O3 are much more volatile than those obtained in the absence of these oxides. More than 60% is a mixture of RuF6 and RuO4 with a smaller amount of RuOF4, RuF4 and RuF5. When the RuO2-Y2O3 or UO2 are dried at 800°C prior to fluorination, RuO2 and RuF6 are also produced, but the yields is about 10∼20%. Results from I. R. measurements are interpreted the same way.

Application of Extraction Chromatography to Nuclear Fuel Reprocessing

The potentialities of applying extraction chromatography to the reprocessing of reactor fuels on an industrial scale have been investigated. The stationary phase was undiluted (100%) tri-n-butyl phosphate (TBP) and the mobile phases were nitric acid or nitrate salt solutions with or without reducing agents for plutonium. Several extraction chromatographic processes for the recovery of nuclear grade uranium and plutonium are described. The flowsheets are based on a systematic determination of the distribution coefficients of relevant metal species (particularly those of uranium, neptunium, plutonium, americium, ruthenium, zirconium and niobium) in the chromatographic systems employed. The Purochromex process developed for the recovery of uranium and plutonium from light-water reactor fuels and the Eurochromex process developed for the separation of highly enriched uranium from irradiated U/Al alloy, U/Zr alloy and uranyl sulfate fuels have successfully been hot-tested on a laboratory scale and cold-tested on an industrial scale. Some complementary studies related to the separation processes, such as radiation degradation of the stationary phase and the removal of tributyl phosphate from product and waste streams, are also described.

Ruthenium in fuel reprocessing by fluoride volatility process. II. Ruthenium species appearing in fluorination step as identified by thermochromatography and IR measurements.

The following three reactions were studied by thermochromatography in relation to the reprocessing of oxide fuels: (a) fluorination of RuF3 by F2, (b) fluorination of RuO2 by F2, (C) fluorination of RuO2 mixed with an excess of an oxide UO2 or Y2O3. The reaction products were as follows. The RuF3-F2 reaction yields RuF5. Fluorination of RuO2 by F2 results in the production of a mixture of RuF4-RuF5-RuOF4. (RuF5 may be produced by fluorination of RuF4 resulting from decomposition of RuOF4 in the apparatus for thermo-chromatography). When dried at 150°C prior to fluorination the products of RuO2-F2 reaction in the presence of an excess UO2 on Y2O3 are much more volatile than those obtained in the absence of these oxides. More than 60% is a mixture of RuF6 and RuO4 with a smaller amount of RuOF4, RuF4 and RuF5. When the RuO2-Y2O3 or UO2 are dried at 800°C prior to fluorination, RuO2 and RuF6 are also produced, but the yields is about 10∼20%. Results from I. R. measurements are interpreted the same way.

Simulation of nuclear-fuel solvent-extraction reprocessing 7. Separation of macroscopic amounts of plutonium and uranium by displacement reextraction of plutonium in reprocessing fast-reactor fuel (section 1)

These data have been obtained from simulating displacement reextraction in a system with two inputs (system for reprocessing a fast-reactor fuel), and they show that it is possible to obtain a uranium extract with not more than 100 μg Pu/kg U in a counter current system with 18 stages to provide partial separation of the uranium and plutonium by reprocessing of an organic solution containing U+ Pu (10 ∶ 1) with 64% saturation in the sum of the metals to produce reextract containing plutonium with U : Pu≤3 and over 99.99% extraction of the uranium; this requires 90 g/liter of uranium in the reextractant and the parameters n=2.06−2.00 (\(X_{HNO_3 }^0 \)=0.3 M); n=1.82−1.87 (\(X_{HNO_3 }^0 \)=0.5 M); n=1.61−1.78 (\(X_{HNO_3 }^0 \)=1.0 M).

Ruthenium in Fuel Reprocessing by Fluoride Volatility Process

The distribution of fission products in various steps of the fluoride volatility process on simulated irradiated nuclear fuels is considered. The ruthenium behavior is similar to that of Zr. This suggests a large percentage of Ru appears as RuF4 presumably after decomposition of the instable RuOF4 produced in the fluorination in the presence of oxygen.

Fuel Reprocessing Still the Focus of U.S. Nonproliferation Policy

Fuel Reprocessing Still the Focus of U.S. Nonproliferation Policy

Use of Electrochemical Processes in Aqueous Reprocessing of Nuclear Fuels

Use of Electrochemical Processes in Aqueous Reprocessing of Nuclear Fuels

Behavior of Ruthenium in Fluoride-Volatility Processes, (IV)

The products of the following three reactions were studied in relation to the reprocessing of oxide fuels: (i) fluorination of Ru by F2 (ii) fluorination of Ru by a O2-F2 mixture, and (iii) secondary process of RuO2-F2 reaction. The product of Ru-F2 reaction was only RuF5; the mass spectrum of RuF5 was obtained. Fluorination of Ru by a O2-F2 mixture resulted in the production of RuF5 (85∼75%) and RuOF4 (15∼25%); these results are different from those reported by earlier workers. The use of radioactive Ru*O2 traced the behavior of RuOF4 in the apparatus. RuOF4 decomposes on the wall which was not preliminarily coated with ruthenium; refluorination was effective for removal of the deposit. These results suggest that the fluorination of irradiated oxide fuels volatilizes the ruthenium as a mixture of RuOF4, RuF5, and a small amount of RuO4.

International Arrangements for Nuclear Fuel Reprocessing. Edited by Abram Chayes and W. Bennett Lewis. (Cambridge, Mass.: Ballinger Publishing Co., 1977. Pp. xx, 251. Index. $15.)

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Durability of vitrified highly active waste from nuclear reprocessing

THE future of nuclear power is often said to depend on the development of a satisfactory method for dealing with highly radioactive waste such as the liquid which is produced during the extraction of plutonium from irradiated nuclear fuel. (See recommendation 27 in ref. 1). Much attention is now being given to developing processes suitable for solidifying highly active liquid waste2 and conversion to a borosilicate glass seems the most favourable route. Here we report a recent investigation at Harwell which supports the belief that the installation of a satisfactory full-scale vitrification process will be successfully achieved.

Reply to “Comments on ‘Solvent Performance in THTR Nuclear Fuel Reprocessing. Part II: On the Formation of Dibutyl Phosphoric Acid by Radiolytic and Hydrolytic Degradation of the TBP-n-Paraffin Extractant’”

Reply to “Comments on ‘Solvent Performance in THTR Nuclear Fuel Reprocessing. Part II: On the Formation of Dibutyl Phosphoric Acid by Radiolytic and Hydrolytic Degradation of the TBP-<i>n</i>-Paraffin Extractant’”

Comments on “Solvent Performance in THTR Nuclear Fuel Reprocessing. Part II: On the Formation of Dibutyl Phosphoric Acid by Radiolytic and Hydrolytic Degradation of the TBP-n-Paraffin Extractant”

Comments on “Solvent Performance in THTR Nuclear Fuel Reprocessing. Part II: On the Formation of Dibutyl Phosphoric Acid by Radiolytic and Hydrolytic Degradation of the TBP-<i>n</i>-Paraffin Extractant”

New prospects for low cost thorium cycles

The rapid increase in the price of uranium and recent increases in the cost of uranium enrichment and prospective costs of fuel reprocessing suggest that long irradiations of fuel and the post-ponement of fuel reprocessing would give the lowest cost fuel cycles. The flexibility of using two separate fuels appears to leave the cost advantage with such cycles avoiding any reprocessing of plutonium-bearing fuels and extraction of uranium-233 from thorium fuels only after several years of irradiation. The resulting fuel cycle costs in prospect can be low and very attractive. The analysis employs the LATREP code.

Behavior of Ruthenium in Fluoride-Volatility Processes, (IV)

The products of the following three reactions were studied in relation to the reprocessing of oxide fuels: (i) fluorination of Ru by F2 (ii) fluorination of Ru by a O2-F2 mixture, and (iii) secondary process of RuO2-F2 reaction. The product of Ru-F2 reaction was only RuF5; the mass spectrum of RuF5 was obtained. Fluorination of Ru by a O2-F2 mixture resulted in the production of RuF5 (85∼75%) and RuOF4 (15∼25%); these results are different from those reported by earlier workers. The use of radioactive Ru*O2 traced the behavior of RuOF4 in the apparatus. RuOF4 decomposes on the wall which was not preliminarily coated with ruthenium; refluorination was effective for removal of the deposit. These results suggest that the fluorination of irradiated oxide fuels volatilizes the ruthenium as a mixture of RuOF4, RuF5, and a small amount of RuO4.

The Use of Fuel Reprocessing Plant Instrumentation for International Safeguards

The International Atomic Energy Agency has a program for developing instrumentation to be used by safeguards inspectors at reprocessing facilities. These instruments have generally been individual pieces of equipment for improving the accuracy of existing measurement instrumentation or equipment to perform nondestructive assay on a selected basis. It is proposed that greater use be made of redundant plant instrumentation and data recovery systems that could augment plant instrumentation to verify the validity of plant measurements. Use of these methods for verification must be proven as part of an operating plant before they can be relied upon for safeguards surveillance. Inspectors must be qualified in plant operations, or have ready access to those so qualified, if the integrity of the operation is to be properly assessed. There is an immediate need for the development and in-plant proof testing of an integrated gamma, passive neutron, and active neutron measurement system for drum quantities of radioactive trash. The primary safeguards effort should be limited to plutonium and highly enriched uranium.

Solvent Performance in THTR Nuclear Fuel Reprocessing. Part II: On the Formation of Dibutyl Phosphoric Acid by Radiolytic and Hydrolytic Degradation of the TBP-n-Paraffin Extractant

Radiolysis and hydrolysis of tributylphosphate (TBP) in n-paraffin diluent were investigated under conditions simulating several parameters of the THOREX flowsheet for reprocessing thorium containing spent nuclear fuel. The solvent (30 vol% TBP-n-paraffin) was equilibrated with 0.25 to 3 molar nitric acid, and the organic phase was inspected by gas chromatography. Radiolysis in pure solvents (30 and 5 vol% TBP-n-paraffin) with doses from 0.5 to 13 Wh . l/sup -1/ produces HDBP with G values (TBP-based) of approx. 2 and approx. 6, slightly decreasing with higher doses. No H/sub 2/MBP was found. Radiolysis in solvent (30 vol% TBP-n-paraffin), equilibrated with 0.25 to 3 molar nitric acid at a constant dose of 7 Wh . l/sup -1/, leads to a considerably reduced radiolytic fraction of the total HDBP yield with a minimum of approx. 0.3 molar nitric acid in the organic phase (2 molar in the aqueous phase). Irradiating the solvent (30 vol% TBP-n-paraffin), equilibrated with 0.75 molar nitric acid (0.12 molar in the organic phase) with doses from 0.5 to 13 Wh . l/sup -1/, leads to a mixture of hydrolytically and radiolytically formed HDBP, the hydrolytic fraction of which raises the total HDBP yield over the radiolytic HDBP yield frommore » acid-free solvents up to doses of approx. 3 Wh . l/sup -1/. It is proposed to denominate TBP degradation product yields according to their origin, hydrolysis or radiolysis (G value), or just by a concentration value in case of mixed origin, valid for the experimental conditions.« less