Yoshikazu Koma

Partitioning of Cesium from a Simulated High Level Liquid Waste by Extraction Chromatography Utilizing a Macroporous Silica‐Based Supramolecular Calix[4]arene‐Crown Impregnated Polymeric Composite

Abstract 1,3‐[(2,4‐Diethyl‐heptylethoxy)oxy]‐2,4‐crown‐6‐calix[4]arene (Calix[4]arene‐R14) is a supramolecular compound exhibiting Cs ion recognition in high level liquid waste (HLLW). To separate effectively Cs(I) from HLLW, a novel silica‐based Calix[4]arene‐R14 polymeric material (Calix[4]arene‐R14/SiO2‐P) was prepared. It was done through impregnating a mixture of Calix[4]arene‐R14 and tri‐n‐butyl phosphate (TBP), a molecular modifier, into the pores of macroporous SiO2‐P particles utilizing a vacuum sucking technique. The sorption of some typical fission product (FP) elements Na(I), K(I), Cs(I), Rb(I), Sr(II), and La(III) towards Calix[4]arene‐R14/SiO2‐P was investigated by examining the influence of contact time and the HNO3 concentration. It was found that with an increase in the HNO3 concentration from 1.0 M to 4.0, the sorption of Cs(I) towards Calix[4]arene‐R14/SiO2‐P increased quickly and then decreased with further increase of HNO3 concentration to 5.0 M. At the optimum HNO3 concentration of 4.0 M HNO3, the Calix[4]arene‐R14/SiO2‐P adsorbent exhibited excellent sorption ability and selectivity for Cs(I) over all of the tested elements, which showed very weak or almost no sorption except Rb(I). The chromatographic separation of Cs(I) from a simulated HLLW containing ∼5 mM of the FP elements and 4.0 M HNO3 was performed by Calix[4]arene‐R14/SiO2‐P packed column at 298 K. Na(I), K(I), Sr(II), and La(III) were found to elute facilely and flowed into the effluent along with 4.0 M HNO3. Cs(I) and Rb(I), adsorbed strongly onto Calix[4]arene‐R14/SiO2‐P, were desorbed sufficiently by water. The leakage behavior of Calix[4]arene‐R14 and TBP from Calix[4]arene‐R14/SiO2‐P was also investigated.

Conceptual Design Study on Advanced Aqueous Reprocessing System for Fast Reactor Fuel Cycle

The design study of an aqueous reprocessing system has been progressed for the feasibility study on commercialized fast reactor cycle systems in Japan. A simplified PUREX process, with the addition of a uranium crystallization step and a minor actinide (MA) recovery process was proposed as the NEXT process. For the simplified PUREX process and the SETFICS/TRUEX process for MA recovery, small-scale hot tests were conducted. The test results showed the decontamination factor (DF) of 105 for the U, Pu and Np product was a reasonable target for the simplified PUREX process. Concerning to the Am and Cm product, the DF of rare earth elements (REs) and the other fission products (FPs) were over 10 and over 102, respectively. For economical competitiveness, an aqueous reprocessing plant with the capacity of 200 tHM/yr cleared the target cost, while the plant with the capacity of 50tHM/yr did not clear it. For efficient utilization of resources, the calculated U and TRU recovery of both plants was over 99wt%. Using salt-free reagents and repeating concentration of waste solutions, the environmental impact was reduced. Pu was not isolated through the whole processes in the NEXT process in order to keep the proliferation resistance.

Separation of Actinide Elements by Solvent Extraction Using Centrifugal Contactors in the NEXT Process

Using the advanced aqueous reprocessing system named NEXT process, actinides recovery was attempted by both a simplified solvent extraction process using TBP as an extractant for U, Pu and Np co-recovery and the SETFICS process for Am and Cm recovery from the raffinate. In U, Pu and Np co-recovery experiments a single cycle flow sheet was used under high nitric acid concentration in the feed solution or scrubbing solution. High nitric acid concentration in the feed solution aided Np oxidation not only in the feed solution, but also at the extraction section. This oxidation reaction accomplished Np extraction by TBP with U and Pu. Most of Np could be recovered into the product solution. In the SETFICS process, a TRUEX solvent of 0.2 mol/dm3 CMPO and 1.4 mol/dm3 TBP in n-dodecane was employed instead of 0.2 mol/dm3 CMPO and 1.0 mol/dm3 TBP in n-dodecane in order to increase the loading of metals. Instead of sodium nitrate, hydroxylamine nitrate was applied to this experimental flow sheet in accordance with a “salt-free” concept. The counter current experiment succeeded with the Am and Cm product. On the high-loading flow sheet, compared with the previous flow sheet, the flow of the aqueous effluents and spent solvent were expected to decrease by about one half. Two solvent extraction experiments for actinides recovery demonstrated the utility of the flow sheet of these processes in the NEXT process.

Enhancement of the mutual separation of lanthanide elements in the solvent extraction based on the CMPO-TBP mixed solvent by using a DTPA-nitrate solution

The mutual separation behavior of rare earth elements was studied in the system based on the neutral extractant CMPO (n-octyl(phenyl)-N,N-diisobutylcarbamoyl methylphosphine oxide) and aminopolyacetic acid DTPA (diethylenetriamine pentaacetic acid). The extractability of lanthanides decreased monotonously with increasing of atomic number in the system of 0.2 M CMPO-1.0 M TBP-n-dodecane (TRUEX solvent) and 0.05 M DTPA-NaNCO3 or hydroxylamine nitrate (HAN) solution. The separation factor of La/Lu was 6,600 using 0.05 M DTPA-3M NaNO3 solution (initial pH=2.0). The mutual separation of lanthanides in the CMPO/DTPA system was greatly improved when compared with the general TRUEX system. The separation factor was not affected by the initial pH of the solution (1.8–2.2), the salting out reagent (NaNO3 and HAN), or its concentration (2, 3 M). The experimentally obtained separation factors of lanthanide elements were reproduced by a simple model based on the distribution ratios in the TRUEX extraction system and t...

Separation of Trivalent Minor Actinides from Fission Products Using Single R-BTP Column Extraction Chromatography

As part of the Fast Reactor Cycle Technology Development (FaCT) Project, research and development has been underway on a system for reprocessing spent fuel from fast breeder reactors. In this system, the method of extraction chromatography is used to recover minor trivalent actinides (MA(III) = Am(III) + Cm(III)) from acidic raffinate in the solvent extraction process that recovers U, Np, and Pu. In general, extractants for solvent extraction can be used as impregnated adsorbents for the extraction chromatography system. The principle of separation used in extraction chromatography is similar to that of solvent extraction. Because of the similarity in chemical properties between MA(III) and Lanthanides(III), MA(III) recovery processes using solvent extraction consist of two steps, namely, MA(III) Ln(III) recovery and MA(III)/Ln(III) separation. Of these, MA(III)/ Ln(III) separation is one of the most challenging issues. Since nitrogen and sulfur donors bind more readily to MA(III) than to Ln(III), a large number of N-donor extractants have been developed in many research projects. Kolarik et al. have reported that a new N-donor ligand, 2,6-bis(5,6-dialkyl-1,2,4-triazine-3-yl)pyridine (R-BTP), shows high selectivity for MA(III) over Ln(III). To develop a partitioning process using the extraction chromatography technique, Wei et al. reported an excellent adsorbent for extraction chromatography. The support particle for the adsorbent consisted of porous silica supports coated with styrenedivinylbenzene polymer (SiO2-P). In many partitioning methods using extraction chromatography, MA(III) separation from the raffinate is achieved using a two-column unit system corresponding to MA(III) Ln(III) recovery and MA(III)/Ln(III) separation, which is similar to the solvent extraction process. In this study, the authors attempted MA(III) direct separation from simulated raffinate using a novel single-column unit system. The partitioning plant for the fast reactor cycle using the singlecolumn unit system will be a simple and compact structure compared with the two-column unit system. The column for the single-column system was packed with the adsorbent, which was impregnated with R-BTP into SiO2-P, for the separation of MA(III) from the raffinate. The behaviors of Am(III) and Cm(III) in separating from the raffinate were examined through a column experiment using a single column packed with the R-BTP/SiO2-P adsorbent.

Leakage of Octyl(Phenyl)-N,N-Di-Isobutylcarbamoylmethylphosphine Oxide from a Macroporous Silica-Based Chelating Polymeric Adsorption Material and its Recovery by Some Selected Porous Adsorbents

The leakage behaviour of a neutral chelating agent, i.e. octyl(phenyl)-N,N-di-isobutylcarbamoylmethylphosphine oxide (CMPO), from its macroporous silica-based (CMPO/SiO2-P) extraction resin (a novel polymeric adsorption material developed in the MAREC process) and the recovery of CMPO by some selected polymer- and silica-based porous adsorbents were investigated by column operation in 0.01 M HNO3 at 25°C or 50°C. The concentration of CMPO in the effluent leaking from CMPO/SiO2-P was determined as ca. 48 ppm at 25°C and ca. 37 ppm at 50°C. The corresponding elution volumes and the amounts of CMPO leaking were 8700.9 BV/g and 239.3 mg/g at 25°C and 11 842.9 BV/g and 359.4 mg/g at 50°C, respectively. The adsorption experiment showed that a polymer-based porous material, SEPABEADS® SP-825, was capable of recovering CMPO effectively from 0.01 M HNO3. Its adsorption ability towards CMPO was considerably greater than those of macroporous silica-based SiO2-P particles, polymer-based Amberlite™ XAD-7 and activated carbon. The bed volume and the amount of CMPO adsorbed by SEPABEADS® SP-825 at the breakthrough point at 25°C were 13 549.2 BV/g and 498.7 mg/g, respectively, thereby demonstrating that SEPABEADS® SP-825 was promising for the recovery of CMPO from Am(III)- and Cm(III)-containing solutions in the MAREC process.

Estimation of the inventory of the radioactive wastes in Fukushima Daiichi NPS with a radionuclide transport model in the contaminated water

ABSTRACT Quantification of the radioactive waste inventory remaining inside the reactors at Fukushima Daiichi NPS is necessary to effectively plan their recovery, treatment, and disposal. Analysis of radionuclide concentrations and secondary wastes in the contaminated water treatment system can provide a means to estimate the radioactive waste inventory, which is not possible by more direct methods due to problems of accessibility and high levels of radiation. A predictive model has therefore been developed to estimate the radioactive waste inventory from the radionuclide concentrations and throughputs in the contaminated water. Model fitting has enabled the estimation of the key parameters, such as the initial radionuclide concentration C0, the continuous release rate F, and inventory of source of continuous release IS0. An estimated one-third of the total 137Cs inventory has already made its way into the water treatment system as secondary wastes, whereas half still remains inside the damaged reactors as of 13 March 2014.

Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO3 concentrations. The flowsheet with a supply of 0.15 mol/dm3 HNO3 solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×105 and 2.42×105, respectively.

Effect of Decontamination Factor on Core Neutronic Design of Light Water Reactors Using Recovered Uranium Reprocessed by Advanced Aqueous Method

In the case where uranium recovered by an advanced aqueous reprocessing is utilized in light water reactors (LWRs) with the thermal neutron spectrum, the effects of the decontamination factor (DF) of the reprocessing on core neutronic characteristics were examined. The amounts of transuranium (TRU) elements and fission products (FPs) contained in the recovered uranium depend on the DF of reprocessing, and also 236U is generated by neutron capture of 235U during a reactor operation. These all act as poisons in the fuel. Therefore, in this paper, the additional 235U enrichment necessary to compensate for the produced TRU elements, FPs, and 236U was evaluated for three representative DF values: 102, 103, and infinity. The low value of 102 corresponds to the advanced aqueous reprocessing investigated here. An APWR core with a discharge burnup of 49 GWd/t when the initial 235U enrichment is 4.6% was considered as the reference core. Uranium of the reference core was enriched from the natural one. It was calculated that the recovered uranium has to be re-enriched up to 5.24% even when DF is infinity in order to achieve the same burnup of 49GWd/t as the reference core. On the other hand, it was also found that the necessary 235U enrichment after the advanced aqueous reprocessing studied here with the low DF value of 102 is only slightly different. The effect of the DF value on moderator reactivity coefficient was also studied, and no effect was found.

Chromatography Column System With Controlled Flow and Temperature for Engineering Scale Application

Japan Atomic Energy Agency (JAEA) is conducting R&D of the engineering scale extraction chromatography system, which uses silica-based adsorbents impregnated with an extractant for the minor actinides (Am and Cm) recovery from the high level liquid waste generated in the spent FBR fuel reprocessing, as a part of the Fast Reactor Cycle Technology Development (FaCT) project. A bench scale testing system was made and provided for the first step of development. The column of the test system (ID 480 or 200 mmΦ with 650 mm height) was equipped with ports for 6 external sensors at its top, middle and bottom levels for measuring the flow velocity or temperature, and for additional 6 heaters for simulating the decay heat of Am and Cm at the middle level of the column. The flow velocity distribution was almost constant except for the very near at the column wall, and it was almost uniform when the liquid flew from top to bottom direction with 4 cm/min of the velocity. The heaters scarcely influenced on the temperature profile inside the column when the power applied to the heater simulated the decay heat of Am, Cm and FPs. The decay heat generated in the column was transported to the effluents and the temperature inside column was kept almost constant.Copyright