Novel imino diacetamide grafted styrene divinyl benzene resin for separation and recovery of palladium from simulated high level liquid waste

Imino diacetamide styrene divinyl benzene resin was evaluated for separation of palladium from simulated high level liquid waste (HLLW). The kinetics of sorption was found to be fast, and the kinetic data were fitted well to the pseudo-second-order kinetics model. Very high K_d (~10³) were obtained for Pd for feed solutions having acidity from 0.25 M to 4.0 M HNO₃. The loading studies of the resin showed the maximum sorption capacity of 54 mg/g. Back extraction studies showed that sorbed Pd can be effectively back extracted using 0.01 M thiourea in 0.2 M HNO₃. Studies with simulated HLLW showed preferential sorption of Pd over other metal ions.     

Impregnation synthesis of a novel macroporous silica-based crown ether polymeric material modified by 1-dodecanol and its adsorption for strontium and some coexistent metals

4,4′,(5′)-Di-(tert-butylcyclohexano)-18-crown-6(DtBuCH18C6) is a chelating agent having high selectivity mostly for Sr(II). To significantly reduce its leakage by molecular modification, a macroporous silica-based DtBuCH18C6 polymeric composite (DtDo/SiO₂–P) was synthesized. It was performed by impregnating and immobilizing DtBuCH18C6 and 1-dodecanol molecules into the pores of the SiO₂–P particles utilizing an advanced vacuum sucking technique. The adsorption of a few fission and non-fission products Sr(II), Ba(II), Cs(I), Ru(III), Mo(VI), Na(I), K(I), Pd(II), La(III), and Y(III) onto DtDo/SiO₂–P was investigated. It was done by examining the effects of contact time and the HNO₃ concentration in a range of 0.1–5.0 M at 298 K. At the optimum concentration of 2.0 M HNO₃, DtDo/SiO₂–P exhibited strong adsorption ability and high selectivity for Sr(II) great over all of the tested elements, which showed very weak or almost no adsorption except Ba(II). Meanwhile, It was found that the quantity of total organic carbon (TOC) leaked from DtDo/SiO₂–P in 2.0 M HNO₃, 187.5 ppm, was lower than 658.4 ppm that leaked from DtBuCH18C6/SiO₂–P, which was not modified. This was ascribed to the effective association of DtBuCH18C6 and 1-dodecanol through intermolecular interaction. The reduction of DtBuCH18C6 leakage by molecular modification with 1-dodecanol was achieved. It was of great benefit to application of DtDo/SiO₂–P in chromatographic partitioning of Sr(II), one of the main heat generators, from high level liquid waste (HLLW) in reprocessing of nuclear spent fuel in the MAREC (Minor Actinides Recovery from HLLW by Extraction Chromatography) process developed recently.     

Studies on the Development of a Flow-Sheet for AHWR Spent Fuel Reprocessing Using TBP

The present paper describes the results of solvent extraction studies carried out in batch mode to collect data on distribution of uranium, plutonium, and thorium using 5% TBP in n-dodecane. Extraction studies are carried out from feed solutions having bulk thorium containing aluminum and fluoride ions in ～3.00–4.00 M nitric acid at concentration levels anticipated in feed solutions during Advanced Heavy Water Reactor (AHWR) spent fuel reprocessing. Studies are carried out under varied experimental conditions. Parameters such as organic to aqueous phase ratio during extraction, concentration of nitric acid for scrubbing co-extracted thorium from loaded organic phase etc., are studied in detail. Hydroxylamine nitrate is selected for reductive stripping of plutonium in preliminary studies. Reagent mixture containing 0.30 M HAN + 0.60 M HNO₃ and 0.20 M N₂H₄ is found to be optimum for plutonium partitioning. This paper also describes the extraction and stripping of uranium and plutonium in co-current mode. The extraction behavior of relevant fission products is studied from a simulated feed solution. A preliminary study on a few commercially available reducing agents is also included. These data are useful in developing a flow-scheme for the recovery of uranium and plutonium from spent fuel originating from AHWR.     

Recovery of Fission Product Palladium from Acidic High Level Waste Solutions

Abstract

The recovery of palladium from a synthetic pressurized heavy water reactor high level waste (PHWR-HLW) solution has been carried out, and the best reagents to use for the actual HLW solutions are discussed. The extraction of palladium from nitric acid solutions has been carried out using Cyanex-471X (triisobutylphosphine sulfide, TIPS) as the extractant. The metal ion could be quantitatively extracted from solutions with nitric acid concentrations between 2.0 and 6.0 M. The species extracted into the organic phase was found to be Pd(NO₃)₂ TIPS. Nitric acid in the range of 2. 0 to 5. 0 M had no effect on TIPS for at least 71 hours. A systematic study of gamma irradiation on loading and stripping of palladium from loaded organic phases using several potential extractants, TIPS, alpha benzoin oxime, dioctylsulfide, and dioctylsulfoxide has been made. A flow sheet for the recovery of palladium from actual HLW solutions using TIPS is proposed.     

Studies of sorption of Pd(II) microquantities on strongly basic polyacrylate anion exchangers

The sorption of palladium(II) microquantities from a chloride and chloride-nitrate (V) model 0.1-6.0 M HCl, 0.1-0.9 M HCl-0.9-0.1 M HNO₃, 1.0 M ZnCl₂-0.1 M HCl and 1.0 M AlCl₃-0.1 M HCl systems on strongly basic polyacrylate anion exchangers was studied. The research was carried out by means of dynamic and static methods. The recovery factors of Pd(II) were determined from the static studies. The working ion-exchange capacities as well as weight and bed distribution coefficients of Pd(II) were calculated from the breakthrough curves. The commercial strongly basic anion exchangers of the functional quaternary ammonium groups — Purolite A-850 and Amberlite IRA-958 — can be used in recovery technologies of Pd(II) from anodic slime, wastewaters and spent catalysts due to their high selectivity.     

Radioactive Tracer Tests of Two Improved TRPO Processes for High-Level Liquid Waste Using Annular Centrifugal Contactors

High-level liquid waste (HLLW) produced from the reprocessing of the spent nuclear fuel still contains moderate amounts of uranium, transuranium (TRU) actinides, and fission products, and thus constitutes a permanent hazard to the environment. The partitioning and transmutation (P&T) strategy has increasingly attracted interest for the safe treatment and disposal of HLLW, in which the partitioning of HLLW is one of the critical technical issues. Two improved trialkylphosphine oxide (TRPO) processes for the removal of actinides have been developed to treat Chinese HLLW, based on the original TRPO process. In one improved process N,N-diethylhydroxylamine as a reducing agent was used for reducing Np(V) and Np(VI) to Np(IV) in order to improve the extraction efficiency of Np. In the other improved process, ammonium vanadate as an oxidizing agent was used for oxidizing Np(V) and Np(IV) to Np(VI) in order to improve the extraction efficiency of Np. Radioactive tracer tests of two improved TRPO processes were carried out using 30-stage 10-mm-diam annular centrifugal contactors and simulated HLLW containing U, Np, Pu, and Am. The test results showed that the decontamination factor of total α activity was >1 × 10⁵. During the test, 30-stage 10-mm-diam annular centrifugal contactors worked in a stable manner continuously, with no stage failing or any interruption of the operation.     

Sorption of palladium on carbon adsorbents from nitric acid solutions

Sorption recovery of palladium from nitric acid solutions on carbon adsorbents BAU, LKAU-7, ABG and UC has been investigated using model solutions with concentrations 8 × 10⁻⁴–8 × 10⁻³ mol/l for palladium and 1, 2 and 5 mol/l for nitric acid. The recovery degrees of Pd(II) depend on the concentration of palladium in contacting solutions as well as on the type of sorbent used. On average, they reach 60%–100% with the maximum in 1 M HNO₃ The palladium desorption by 10% thiocarbamide solution in 1M H₂SO₄ proceeds completely for the sorbent LKAU-7. The use of thiocarbamide solutions in 0.1 M NaOH increases the desorption of palladium from the sorbents BAU and UC up to 80%–85%     

Studies on the Separation and Recovery of Palladium from Simulated High Level Liquid Waste (SHLW) Solution with Novel Extractant N,N,N′,N′-tetra (2-ethylhexyl) Dithiodiglycolamide DTDGA

The extraction properties of N,N,N',N'-tetra-(2-ethylhexyl) dithiodiglycolamide (DTDGA) have been evaluated for separation and recovery of palladium from simulated high-level liquid waste (SHLW) solution. Extraction of Pd was found to increase with increase in nitric acid concentration up to 4.0 M, above which the extraction remains almost constant. Acid uptake studies show 1:1 stoichiometry between DTDGA and HNO₃ at nitric acid concentration above 3.0 M. The acid uptake constant (K_H) was found to be 0.60 which could be due to the presence of two carbonyl groups of amidic moiety. DTDGA has shown very high extractability and selectivity for Pd over other metal ions present in SHLW. The separation factor (SF) for Pd over other metal ions was found to be more than 10⁴. Almost complete back extraction of palladium from organic phase was achieved with 0.01 M thiourea in 0.1 M nitric acid. Reusability studies of the extractant indicate that D_Pd remains almost constant even after five successive cycles of extraction and stripping. Two extraction stages will be required for complete extraction of 100 mg/L Palladium in 3.0 M nitric acid solution using 0.0025 M DTDGA/n-dodecane solvent system.     

Extraction–electrodeposition (EX–EL) process for the recovery of palladium from high-level liquid waste

Extraction–electrodeposition (EX–EL) process has been developed for the quantitative recovery of palladium from nitric acid medium and fast reactor-simulated high-level liquid waste (FR-S-HLLW). The process exploits some characteristic properties of room temperature ionic liquid, tri-n-octylmethylammonium nitrate (TOMAN), for quantitative and favorable recovery of palladium. Extraction of palladium (II) from FR-S-HLLW and nitric acid medium by a solution of 0.5 M TOMAN in chloroform has been studied in detail. More than 60% of palladium was extracted in a single contact of equal volumes of organic and aqueous phases and nearly five contacts were required for quantitative extraction. The electrochemical behavior of palladium (II) present in the organic phase was investigated at stainless steel electrode by cyclic voltammetry. A surge in cathodic current occurring at a potential of –0.5 V (vs. Pd) was due to the reduction of palladium (II) to palladium (0). The kinetics of electrodeposition was followed by the UV–VIS absorption spectrum of palladium present in organic phase and under the given conditions nearly 20 and 35 h were required for the quantitative deposition of palladium from organic phase, which was obtained after extraction of palladium from 4 M nitric acid and FR-S-HLLW, respectively. Decontamination of palladium from other fission products during extraction and electrodeposition was studied and the results are reported in this article.     

Precipitation characteristics of palladium from a simulated radioactive liquid waste by Ascorbic Acid

The precipitation characteristics of palladium (Pd) from a simulated radwaste solution by ascorbic acid was investigated. Pd was selectively precipitated by the reduction reaction of Pd with ascorbic acid. When the nitric acid concentration in the 8 component system was below 2.0 M, Pd over 99.5 % was selectively precipitated at an ascorbic acid concentration above 0.06 M, while other elements mainly remained in solution. The precipitation yield of Pd by ascorbic acid decreased with the increase of the nitric acid concentration.     

Selective Actinide Extraction with a Tri‐Synergistic Mixture Using a Centrifugal Contactor Battery

Abstract

The reprocessing of spent nuclear fuels produces a quantity of high‐level liquid waste (HLLW), which is subsequentely vitrified. The most important contributors to the long‐term radiotoxicity of the vitrified HLLW are the so‐called minor actinides (MAs), which are neptunium, americium, and curium. This paper presents the results obtained during some extraction tests of americium(III) in a synthetic liquid waste representative of that coming from the PUREX and DIAMEX processes, obtained by dissolving the respective nitrate salts of the lanthanides(III) in a diluted solution of nitric acid and spiked with ²⁴¹Am and ¹⁵²Eu. The extraction process employed a tri‐synergistic organic extractant mixture and a centrifugal contactor battery of eight stages.     

Optimization Studies for the Recovery of Thorium from Advanced Heavy Water Reactor High Level Waste (AHWR-HLW) Solutions Using Green Solvents

An Advanced Heavy Water Reactor (AHWR) has been specifically designed to exploit Th/²³³U as fuels. The reprocessing is focused mainly on the recovery of ²³³U and Pu from the spent fuels leaving bulk of Th (～100 g/L) in the High Level Waste (HLW) solutions. No systematic attempts have been made so far to identify suitable solvents for the recovery of thorium from high level waste (HLW) solutions generated after AHWR spent fuel reprocessing. Tri-n-butyl phosphate (TBP), though the work horse for nuclear fuel reprocessing as an extractant, suffers from the serious limitation of third-phase formation during the extraction of macro concentrations of thorium. Two straight chain dialkyl amides such as N,N-dihexyl octanamide (DHOA), and N,N-dihexyl decanamide (DHDA) as well as TBP were evaluated for the recovery of the thorium from AHWR-HLW solutions. Attempts were made to identify suitable solvent (extractant + diluent) system and optimize the conditions for the recovery of thorium from HLW solutions. Selectivity of the solvents was examined for thorium extraction over fission products/structural materials under AHWR raffinate solutions. Counter-current centrifugal contactor runs were also carried out on simulated waste solutions to validate the optimized conditions for the recovery of thorium from the simulated AHWR waste solutions.     

Separation,Preconcentration, and Recovery of Pd(II) Ions using Newly Modified Silica Gel with Bis(3-Aminopropyl)Amine

Bis(3-aminopropyl)amine bonded silica gel (BAPA-SG) was synthesized and characterized by elemental analysis and FT-IR spectroscopy, and tested for adsorption, preconcentration, and recovery of Pd(II) ions. The effective parameters on the preconcentration of Pd(II) ions such as pH, volume, and flow rate of the Pd(II) solution, and the type and volume of eluent solution, and also matrix ions such as alkaline and heavy metals were investigated. Flame atomic absorption spectrometry was used for the determination of Pd(II) concentration. The modified silica gel could adsorb Pd(II) ions quantitatively from the solutions up to 400 mL at pH 1.0 by the flow rate of 5 mL/min. The retained Pd(II) ions could be easily eluted by using 5 mL of 1% (m/v) thiourea in 1.0 M HCl solution. The recovery of Pd(II) ions was 95 ± 2% at 95% confidence level. The analytical detection limit of Pd was found to be 0.36 µg L^?1 at the preconcentration factor of 80. Selective adsorption of Pd(II) ions over some base metal ions was also investigated. The developed method was applied to spent auto catalyst for palladium recovery, and a certified ore sample for the determination of palladium content.     

Demonstration of an Improved Total Partitioning Process for High Level Liquid Waste Using Annular Centrifugal Contactors

High level liquid waste (HLLW) produced from the reprocessing of the spent nuclear fuel still contains moderate amounts of uranium, transuranium (TRU) actinides, ⁹⁰Sr, ¹³⁷Cs, etc., and thus constitutes a permanent hazard to the environment. The partitioning and transmutation (P&T) strategy has increasingly attracted interest for the safe treatment and disposal of HLLW, in which the partitioning of HLLW is one of the critical technical issues. An improved total partitioning process, including a TRPO (tri-alkylphosphine oxide) process for the removal of actinides, a CESE (crown ether strontium extraction) process for the removal of Sr, and a calixcrown ether extraction process for the removal of Cs, has been developed to treat Chinese HLLW at the Institute of Nuclear and New Energy Technology (INET), Tsinghua University, China. A demonstration test of the improved total partitioning process was carried out using 74-stage 10-mm-dia annular centrifugal contactors and simulated HLLW. The test results showed that the decontamination factors were >1.2 × 10⁶, 4600, and 7500 for Nd, Sr, and Cs, respectively. In the test, Nd was used to simulate Am. During the test, 74-stage 10-mm-dia annular centrifugal contactors worked stable continuously with no stage failing or interruption of the operation.     

Direct Selective Extraction of Actinides (III) from PUREX Raffinate using a Mixture of CyMe4BTBP and TODGA as 1-cycle SANEX Solvent Part III: Demonstration of a Laboratory-Scale Counter-Current Centrifugal Contactor Process

The direct selective separation of the trivalent actinides americium and curium from a simulated Plutonium Uranium Refining by EXtraction (PUREX) raffinate solution by a continuous counter-current solvent extraction process using miniature annular centrifugal contactors was demonstrated on a laboratory scale. In a 32-stage spiked test (12 stages for extraction, 16 stages for scrubbing, and 4 stages for Am/Cm stripping), an extractant mixture of CyMe₄BTBP and TODGA in a TPH/1-octanol mixture was used. The co-extraction of some fission and corrosion product elements, such as zirconium and molybdenum, was prevented by using oxalic acid. Co-extracted palladium was selectively stripped using an L-cysteine scrubbing solution and the trivalent actinides were selectively stripped using a glycolic acid-based stripping solution. It was demonstrated that a selective extraction and high recovery of > 99.4% of the trivalent minor actinides was achieved with low contamination by fission and corrosion products. The product contained 99.8% of the initial americium and 99.4% of the initial curium content. The spent solvent still contained high concentrations of Cu, Cd, and Ni. The experimental steady-state concentration profiles of important solutes were determined and compared with those from computer-code calculations.     

高放废液煅烧物与基础玻璃高温反应物相与结构转变研究

高放废液玻璃固化是我国亟待突破的关键环节之一,而高放废液玻璃固化过程中高放废液与基础玻璃反应发生的物相转变与结构变化一直是关注的重点。本文以模拟高放废液煅烧物和基础玻璃的混合物为研究对象,利用DSC、原位XRD、NMR、Raman等表征方法开展玻璃固化过程中高温反应物相与结构变化研究。研究结果表明:模拟高放废液煅烧物在~700 ℃开始溶解于基础玻璃中,并在800~900 ℃生成系列硅(铝)酸盐中间物相;煅烧物与基础玻璃高温反应过程中,高放废液中的大量玻璃网络改变体溶入基础玻璃中,导致玻璃网络聚合度下降。     

Studies on the Sorption of Palladium using Cross‐Linked Poly (4‐Vinylpyridine‐Divinylbenzene) Resins in Nitric Acid Medium

Various cross‐linked (4, 8, and 12%) gel‐type weak‐base poly(4‐vinylpyridine) (PVP) resins were studied for palladium recovery from nitric acid medium. The sorption of palladium was found to decrease with an increase in cross‐linkage of the resin. 8 and 12% PVP resins exhibited maximum D _Pd(II) values at 2–6 M HNO₃, whereas 4% PVP resin showed maximum D _Pd(II) values at lower acidities (0.1 M HNO₃). FT‐IR, SEM, and XPS techniques were used for the characterization of palladium‐loaded resins. Detailed studies were carried out with the resin of modest cross‐linkage i.e., 8% PVP resin. The sorption isotherm studies revealed that the maximum palladium loading approaches the theoretical capacity of the resin, presuming the sorption of palladium as divalent anion at 4 M HNO₃. The pseudo‐second order kinetics model yielded the best fit for the experimental data of sorption kinetics. An increase in temperature accelerates the rate of palladium extraction and also the addition of chloride ions increases the palladium uptake. Column studies were performed using 4 and 8% PVP resins in 2 and 4 M nitric acid concentrations. The loaded palladium could be eluted efficiently with acidic thiourea solution.     

Ion exchange characteristics of rhodium and ruthenium from a simulated radioactive liquid waste

Radioactive high-level liquid wastes contain significant quantities of platinum group metals (PGM) such as palladium [Pd(II)], rhodium [Rh(III)] and ruthenium [Ru(III)]. The PGM are produced as fission products in a nuclear reactor. In this study, batch and column experiments were carried out to investigate the ion exchange characteristics of Rh(III) and Ru(III) including the effects of the ionic group of ion exchangers, solution temperature, and the concentration of nitric acid by various ion exchangers such as IRN 78 and Dowex 1x 8; the elution characteristics of Rh(III) and Ru(III) by various eluents were also investigated. Anion exchangers such as Dowex 1x 8 with the ionic group of quaternary methyl ammonium had a higher capacity than anion exchanger such as IRN 78 with amine group for the adsorption of Rh(III) and Ru(III) from a simulated liquid waste. The optimal nitric acid concentration for the adsorption of Rh(III) and Ru(III) by anion exchangers was shown to be about 3 M and 2–3 M, respectively.     

钯/活性炭催化剂中贵金属钯的回收

研究了废钯/活性炭催化剂中金属钯的回收。将废钯/活性炭催化剂用高温焙烧的方法除去其中大部分炭,钯渣用甲酸还原后用王水浸出。考察了焙烧、还原及浸出条件对钯回收率的影响,得到的氯化钯回收率95%,纯度99%。     

Separation and Recovery of Uranium,Neptunium, and Plutonium from High Level Waste Using Tributyl Phosphate: Countercurrent Studies with Simulated Waste Solution

ABSTRACT

The present work deals with countercurrent extraction studies on the partitioning of uranium, neptunium, and plutonium using 30% tributyl phosphate (TBP) from simulated high level waste solution generated during reprocessing of spent uranium fuel from pressurized heavy water reactors. The oxidation states of neptunium and plutonium were adjusted either by 0.01 M potassium dichromate or 0.01 M dioxovanadium ion. Neptunium and plutonium, extracted in the TBP phase, were stripped together using a mixture containing 0.05 M ascorbic acid and 0.25 M hydrogen peroxide in 2.0 M nitric acid solution. Although dioxovanadium ion is more effective for proper adjustment of the oxidation states of plutonium and neptunium, subsequent recovery of these actinides from loaded TBP is better if potassium dichromate is used for the valency adjustment. Results of the stagewise analysis of extraction and stripping of actinides using mixer-settlers are presented.