Recovery Of Neptunium Research Articles

Progress towards the Full Recovery of Neptunium in an Advanced PUREX Process

To meet the needs of future closed fuel cycles, the complete recovery of minor actinides, including neptunium, may be required. Neptunium can be fully recovered by modifications to the Plutonium URanium Extraction (PUREX) process but this requires careful control of the Np(V)-(VI) redox reaction in the first solvent extraction contactor to avoid losses to the highly active aqueous raffinate, as occurs in current reprocessing plants. As part of the on-going development of an Advanced PUREX process we report a series of solvent extraction experiments aimed at optimizing neptunium recovery in a process that is based on centrifugal contactors as the extraction equipment. Suitable experimental conditions for Np(V) oxidation were identified through simple stirred 2-phase experiments and single stage mini-centrifugal contactor experiments. A U/Np-active proof-of-principle flowsheet test in a multi-stage centrifugal contactor cascade then demonstrated > 99% extraction of neptunium, thus suggesting the aims for neptunium recovery in advanced fuel cycles can be met by an Advanced PUREX process.

Recovery of Neptunium by Electrolysis of NpN in LiCl-KCl Eutectic Melts

The electrochemical behavior of neptunium nitride, NpN, in the LiCl-KCl eutectic melt containing NpCl3 at 450, 500 and 550°C was investigated from the viewpoint of the application of electrochemical refining in a fused salt to nitride fuel cycle. The electrochemical dissolution of NpN began nearly at the potential theoretically evaluated, though this reaction was irreversible owing to small partial pressure of N2 in the salt and the reaction rate was slow. Under the electrolysis in the NpCl3-LiCl-KCl eutectic melt, NpN was dissolved into the salt as Np3+ at the anode, and Np metal was deposited at the cathode. About 0.5 g of Np metal was obtained by heating the deposit containing the salt at 800°C for 3.6 ks.

Recovery of Neptunium by Electrolysis of NpN in LiCl-KCl Eutectic Melts.

Recovery of Neptunium by Electrolysis of NpN in LiCl-KCl Eutectic Melts.

Neptunium concentration profiles in the Purex process

Numerical simulations of neptunium concentration profiles in the co-decontamination step of the Purex process is carried out in order to understand neptunium extraction behavior and to propose a flow sheet for neptunium recovery from the reprocessing process. A simulation result for a flow sheet containing Pu(VI) in the feed solution shows that Pu(VI) plays a role of scavenger of nitrous acid. The rate constant of neptunium reduction with n-butyraldehyde is evaluated by using the process data. The determined rate constant is available for the numerical simulation of neptunium concentration profiles in the reprocessing process.

Numerical Simulation of Neptunium Extraction Behavior in Co-decontamination Stage for Late Split Purex System

Numerical simulation of neptunium concentration profiles is carried out in order to understand the extraction behavior of neptunium in the co-decontanimation stage which decontaminates through two extraction banks before the uranium and plutonium partitioning steps. Simulation results show that leaking of neptunium to the waste stream at the second extraction bank is caused by reduction from Np(VI) to Np(V) unextracted by TBP in the presence of nitrous acid of 10−3 mol/l. It is found that the prevention of the reduction reaction from Np(VI) to Np(V) is effective for the recovery of neptunium with uranium and plutonium into the product stream. Two methods for the recovery of neptunium are suggested; i) to decrease residence time in the acid adjustment vessel located between first and second extraction banks, ii) to diminish nitrous acid in the aqueous phase by hydrazine. Simulation results show that the recovery of neptunium with uranium and plutonium is improved from 40 to 60% and kom 40 to 70% by the first and the second methods, respectively.

Numerical Simulation of Neptunium Extraction Behavior in Co-decontamination Stage for Late Split Purex System.

Numerical simulation of neptunium concentration profiles is carried out in order to understand the extraction behavior of neptunium in the co-decontanimation stage which decontaminates through two extraction banks before the uranium and plutonium partitioning steps. Simulation results show that leaking of neptunium to the waste stream at the second extraction bank is caused by reduction from Np(VI) to Np(V) unextracted by TBP in the presence of nitrous acid of 10−3 mol/l. It is found that the prevention of the reduction reaction from Np(VI) to Np(V) is effective for the recovery of neptunium with uranium and plutonium into the product stream. Two methods for the recovery of neptunium are suggested; i) to decrease residence time in the acid adjustment vessel located between first and second extraction banks, ii) to diminish nitrous acid in the aqueous phase by hydrazine. Simulation results show that the recovery of neptunium with uranium and plutonium is improved from 40 to 60% and kom 40 to 70% by the fir...

Recovery of Neptunium from Highly Radioactive Waste Solutions of Purex Origin Using Tributyl Phosphate

Abstract The present work deals with the extraction of neptunium into the TBP/dodecane phase under conditions relevant to highly radioactive waste solutions, along with uranium and plutonium, by oxidizing it to the hexavalent state using 0.01 M K2Cr2O7 and subsequently recovering it by selective stripping. Three types of simulated HLW solutions, namely sulfate-bearing (SB, in ∼0.3 M HNO3) and non sulfate wastes originating from the reprocessing of fuels from pressurised heavy water reactors (PHWR) and fast breeder reactors (FBR) (both in 3.0 M HNO3), have been employed in this study. Very high extraction of U(VI), Np(VI), and Pu(VI) was obtained from PHWR and FBR-HLW solutions, whereas extraction was less but reasonably high from the SB-HLW solution. The uptake of cerium at tracer level concentrations in the millimolar range (encountered in HLW solutions) and from the simulated HLW solutions containing 0.01 M K2Cr2O7 by 30% TBP has shown that its extraction takes place only at tracer level concentrations ...

Recovery of neptunium from highly radioactive waste solutions of PUREX origin using CMPO

The partitioning and recovery of237Np from three types of simulated high level waste solutions originating from PUREX processing of spent nuclear fuels such as sulfate bearing high level waste (SB-HLW), HLW from a pressurised heavy water reactor (PHWR-HLW) and from a fast breeder reactor (FBR-HLW) have been carried out using a mixture of 0.2M CMPO and 1.2M TBP in dodecane. Quantitative extraction of neptunium was possible by either oxidizing it to the hexavalent state keeping K2Cr2O7 at 0.01M concentration or by reducing it to tetravalent state keeping Fe2+ at 0.02M concentration. Stripping of neptunium was carried out using different reagents, such as dilute nitric acid, oxalic acid and sodium carbonate. Almost quantitative recovery of neptunium has been achieved during these studies.

Recovery of neptunium from macroamounts of uranium by extraction with 1-phenyl-3-methyl-4-benzoyl-pyrazolone-5

Conditions for neptunium recovery from nitrate solution containing large amounts of uranium by extraction with 1-phenyl-3-methyl-4-benzoyl-pyrazolone-5 (PMBP), and a mixture of PMBP with di-2-ethyl-hexylphosphoric acid (HDEHP) were studied. A two-stage technique, based on removal of uranium macroamounts at the first stage by means of extraction of uranium with PMBP and HDEHP mixture, and neptunium extraction by PMBP at the second stage, was suggested.

Application of Extraction Chromatography to the Recovery of Neptunium, Plutonium and Americium from an Industrial Waste

Abstract A pilot scale investigation was made to evaluate the possible application of the extraction Chromatographie method (LLC) to the partitioning of alpha emitters from liquid wastes containing traces of transuranium elements. A secondary purpose was to obtain pure AmO2, which is used to produce alpha, gamma, and neutron sources. The process developed for “alpha partitioning” consists essentially of the extraction of macro amounts of uranium with 30% TBP in dodecane in mixer-settlers, then coextraction of Np-237, Pu-239, and Am-241 by LLC on a macro column filled with di-n-hexyl-octoxy-methyl-phosphine oxide (POX. 11) adsorbed on an inert support. In each run about 200 liters of initial waste are decontaminated of alpha emitters. The loading step is followed by selective elution of americium, neptunium, and plutonium. The americium eluate is then subjected to the following operations: (1) separation of Am from Fe and Cd by LLC on a TBP column and (2) separation of Am from lanthanide traces by LLC on a...

Neptunium Recovery and Purification at Hanford

From American Chemical Society Production Technology of Np-237 and Pu- 238, Denver, Jan. 1964. Neptunium is routinely recovered from irradiated fuel elements at Hanford's two main separation plants. Initial development tests were started in the Purex plant in 1958, then in the Redox plant in 1959, and recently culminated in the installation of new production systems in both plants for improved recoveries. Both recovery flowsheets employ solvent extraction techniques based on the relative extractability of neptunium-VI. The neptunium is co-extracted with uranium and plutonium in the plants' first extraction cycles and then partitioned and decontaminated in separate neptanium cycles. Excellent decontamination from fission products is achieved without interfering with mamline uranium and plutonium production. Recovered neptunium is purified by anion exchange and shipped offsite for subsequent irradiation to plutonaim-238. Overall separation factors of uranium and fission products from neptunium are greater than 10/sup 7/ and 10/sup 10/ respectively. (auth)

New Neptunium Recovery Facility at the Hanford Purex Plant

From American Chemical Society 146th National Meeting, Denver, Jan. 1964. An additional solvent extraction battery was installed in the Purex Plant for the continuous recovery of Np/sup 237/ from irradiated reactor fuels. A description is given of the recovery equipment. (D.L.C.)

Recent Solvent Extraction Studies at Hanford Laboratories

Some solvent extraction flowsheet development studies performed recently at Hanford Laboratories in support of the Hanford radioactive waste management and neptunium recovery programs are summarized. Flowsheet for the removal of cesium from Purex current waste and stored waste supernatant liquid are discussed. Dipicrylamine-nitrobenzene is used to extract cesium; dilute nitric acid is used to strip cesium from the organic. A one-cycle flowsheet for the removal of strontium and rare earth elements from Purex process waste and separation of these into a strontium and a rare earth fraction is discussed. Extraction of the desired elements from a citrate complexed feed is by di(2-ethylhexyl) phosphoric acid-tributylphosphate-Soltrol solvent. Strontium is removed from the organic by dilute nitric acid in the second column. Rare earths are stripped from the organic by more concentrated nitric acid in a third column. Procedures for recovery of neptunium and plutonium from Purex process acidic waste are described. The solvent is di(2-ethylhexyl phosphoric) acid-Soltrol; Pu(IV) and Np(IV) are extracted from acid solution at such high distribution ratios that adequate recovery is attained in a single batch contact. Studies leading to the flowsheets as well as results of tests of the flowsheets with simulated and actual plant solutions are discussed.