Simulated Fission Products Research Articles

Behavior of U-Zr Alloy Containing Simulated Fission Products during Anodic Dissolution in Molten Chloride Electrolyte

To investigate the distribution of fission product elements in the electrorefining of spent metallic fuel, electrorefining tests were carried out using U-Zr alloy containing simulated fission products. An intermediate region was found between undissolved alloy and Zr-rich regions formed by the anodic dissolution of uranium. The composition of this region determined by wavelength-dispersive X-ray spectroscopy (WDS) analysis corresponded to that of δ-phase (63–78 at% Zr) in the U-Zr binary phase diagram. This region is expected to form in the course of the anodic dissolution sequence that begins from an undissolved U-Zr alloy and ends with a Zr-rich region. From the results of chemical and WDS analyses of anode residue samples, it was predicted that noble metals would stay in the anode residue as long as zirconium remains even after most of the uranium in the alloy was dissolved. Analytical results indicated no large difference betweenthe noble metal behavior in simulated material and in irradiated EBR-II driver fuels. Decontamination factors of uraniumfrom molybdenum and palladium were defined as ratios between their concentrations in the initial alloy loaded into an anode and those in cathode products, and evaluated from the analytical results. Although there was some scattering of data, decontamination factors of around 70 for molybdenum and 10 for palladium would be expected.

Evaluation Study on Properties of a Novel R-BTP Extraction Resin-From a Viewpoint of Simple Separation of Minor Actinides-

In order to develop a simple separation process by extraction chromatography of trivalent minor actinides (MA) and lanthanides (Ln), a novel macroporous silica-based R-BTP (BTP: 2,6–bis–(5,6–dialkyl–1,2,4–triazine–3–yl)–pyridine, R: isohexyl) extraction resin was synthesized by impregnating the R-BTP extractant into the pores of SiO2-P support particles. The adsorption property of simulated fission products including Ln was investigated with HNO3 solution onto the R-BTP extraction resin. It was found that Dy(III), probably similar adsorbability to MA (III), was dominantly adsorbed onto the resin in 2-4 M HNO3 solution at 50°C, compared with the other Ln (III). In addition, the chemical stability and thermal decomposition property of the resin were evaluated and the resin was resistant to HNO3 solution at ambient temperature.

Development of Advanced Reprocessing System Based on Precipitation Method Using Pyrrolidone Derivatives as Precipitants

An advanced reprocessing system for spent FBR fuels based on two precipitation processes has been proposed. In the first process, only U(VI) species is precipitated using a pyrrolidone derivative (NRP) with lower hydrophobicity and donicity, which should yield a lower precipitation ability. In the second process, residual U(VI) and Pu(IV, VI) are precipitated simultaneously using an NRP with higher hydrophobicity and donicity, which should yield a higher precipitation ability. In order to select the precipitants for the first precipitation process, we have examined the precipitation behavior of U(VI), Pu(IV), and Pu(VI) species in HNO3 using N-n-propyl-2-pyrrolidone (NProP), N-n-butyl-2-pyrrolidone (NBP), and N-isobutyl-2-pyrrolidone (NiBP) with lower hydrophobicity and donicity than N-cyclohexyl-2-pyrrolidone (NCP) previously proposed as the precipitant. It was found that NRPs could precipitate U(VI) nearly stoichiometrically and that the decontamination factors for simulated fission products were higher than those in NCP systems. Furthermore, as seen in NCP, it was found that in the U(VI)-Pu(IV) mixtures, a small amount of Pu(IV) was temporarily coprecipitated with U(VI) by NRPs in spite of their lower precipitation ability and then the coprecipitated Pu(IV) component was redissolved with continuous stirring. From these results, NRPs can be proposed as candidate precipitants for the first precipitation process. In particular, NBP is considered to be the most promising precipitant, because of the relatively high solubility of the NProP precipitant, the increases in viscosity of NiBP slurry with stirring, and the relatively fast sedimentation rate of NBP precipitates.

Development of Advanced Reprocessing System Based on Precipitation Method Using Pyrrolidone Derivatives as Precipitants

An advanced reprocessing system for spent FBR fuels based on two precipitation processes has been proposed. In the first process, only U(VI) species is precipitated using a pyrrolidone derivative (NRP) with lower hydrophobicity and donicity, which should yield a lower precipitation ability. In the second process, residual U(VI) and Pu(IV, VI) are precipitated simultaneously using an NRP with higher hydrophobicity and donicity, which should yield a higher precipitation ability. In order to select the precipitants for the first precipitation process, we have examined the precipitation behavior of U(VI), Pu(IV), and Pu(VI) species in HNO3 using N-n-propyl-2-pyrrolidone (NProP), N-n-butyl-2-pyrrolidone (NBP), and N-isobutyl-2-pyrrolidone (NiBP) with lower hydrophobicity and donicity than N-cyclohexyl-2-pyrrolidone (NCP) previously proposed as the precipitant. It was found that NRPs could precipitate U(VI) nearly stoichiometrically and that the decontamination factors for simulated fission products were higher than those in NCP systems. Furthermore, as seen in NCP, it was found that in the U(VI)-Pu(IV) mixtures, a small amount of Pu(IV) was temporarily coprecipitated with U(VI) by NRPs in spite of their lower precipitation ability and then the coprecipitated Pu(IV) component was redissolved with continuous stirring. From these results, NRPs can be proposed as candidate precipitants for the first precipitation process. In particular, NBP is considered to be the most promising precipitant, because of the relatively high solubility of the NProP precipitant, the increases in viscosity of NiBP slurry with stirring, and the relatively fast sedimentation rate of NBP precipitates.

Effect of Americium and Simulated Fission Products Addition on Oxygen Potential of Uranium-Plutonium Mixed Oxide Fuels

AbstractThe oxygen potentials at 1273 K of mixed oxide (MOX) fuels with Am and 26 kinds of fission product elements (FPs), simulating low-decontaminated MOX fuel and high burn-up of up to 250 GWd/t, have been measured by using thermogravimetric analysis (TGA). The oxygen potentials for simulated low-decontaminated MOX fuels were higher than the fuels without FPs and increased with increasing simulated burn-up.

Distribution Coefficients of U(VI), Nitric Acid and FP Elements in Extractions from Concentrated Aqueous Solutions of Nitrates by 30% Tri-n-butylphosphate Solution

The distribution coefficients of uranium, nitric acid and 8 elements simulating fission products were obtained from the concentrated aqueous solutions of nitrates, whose nitrate ion concentrations were 4–15 mol dm−3. The relationships among the distribution coefficient of uranium, the nitrate ion concentration in the aqueous phase, and the concentration of tri-n-butylphosphate uncombined with nitrates in the organic phase were obtained from the data.

Distribution Coefficients of U(VI), Nitric Acid and FP Elements in Extractions from Concentrated Aqueous Solutions of Nitrates by 30% Tri-n-butylphosphate Solution

The distribution coefficients of uranium, nitric acid and 8 elements simulating fission products were obtained from the concentrated aqueous solutions of nitrates, whose nitrate ion concentrations were 4–15 mol dm−3. The relationships among the distribution coefficient of uranium, the nitrate ion concentration in the aqueous phase, and the concentration of tri-n-butylphosphate uncombined with nitrates in the organic phase were obtained from the data.

Effect of Pulse Electrolysis on Morphology of Co-Deposited MOX Granules

In the oxide electrowinning method of nuclear fuel recycling, the applicability of the pulse electrolysis methodto the mixed-oxide (MOX) codeposition process was evaluated. Several experiments were conducted involving uranium, plutonium, and elements simulating fission products (FPs) or corrosion products (CPs). Through these experiments, the effects of impurities and electrolysis pulse waveforms on the deposit were clarified. In particular, the difference between the dissolution kinetics of UO2 and PUO2 during the period in which current was stopped was found to be the most important factor influencing the deposition behaviors. As a result, pulse electrolysis conditions, which could restrain Pu enrichment of the deposited MOX, were confirmed. Furthermore, it was found that the elemental distribution in the MOX granule obtained by pulse electrolysis was homogenized, which is well suited for nuclear fuel fabrication. Finally, a qualitative model of the pulse electrolysis reaction near the field of the cathode surface is proposed.

모의 사용후 핵연료를 이용한 질화물 핵연료 소결체 제조

In order to investigate a nitriding process of spent oxide fuel and the subsequent change in thermal properties after nitriding, simulated spent fuel powder was converted into a nitride pellet with simulated fission product elements through a carbothermic reduction process. Nitriding rate of simulated spent fuel was decreased with increasing of the amount of fission products. Contents of Ba and Sr in simulated spent fuel were decreased after the carbothermic reduction process. The thermal conductivity of the nitride pellet was decreased by an addition of fission product element but was higher than that of the oxide fuel containing fission product elements.

Development of the chromatographic partitioning of cesium and strontium utilizing two macroporous silica-based calix[4]arene-crown and amide impregnated polymeric composites: PREC partitioning process

To partition effectively Cs(I) and Sr(II), two harmful heat emitting nuclides, from a highly active liquid waste by extraction chromatography, two kinds of macroporous silica-based polymeric materials, Calix[4]arene-R14/SiO 2-P and TODGA/SiO 2-P, were synthesized. Two chelating agents, 1,3-[(2,4-diethyl-heptylethoxy)oxy]-2,4-crown-6-calix[4]arene (Calix[4]arene-R14), an excellent supramolecular compound having molecular recognition ability for Cs(I), and N, N, N′, N′-tetraoctyl-3-oxapentane-1,5-diamide (TODGA) were impregnated and immobilized into the pores of SiO 2-P particles support by a vacuum sucking technique. The loading and elution of 11 typical simulated fission and non-fission products from 4.0 M or 2.0 M HNO 3 were performed at 298 K. It was found that in the first column packed with the Calix[4]arene-R14/SiO 2-P, all of the simulated elements were separated effectively into two groups: (1) Na(I), K(I), Sr(II), Fe(III), Ba(II), Ru(III), Pd(II), Zr(IV), and Mo(VI) (noted as Sr-group); (2) Cs(I)–Rb(I) (Cs-group) by eluting with 4.0 M HNO 3 and distilled water, respectively. The harmful element Cs(I) flowed into the second group along with Rb(I) because of their close sorption and elution properties towards Calix[4]arene-R14/SiO 2-P, while Sr(II) showed no sorption and flowed into Sr-containing group. In the second column packed with TODGA/SiO 2-P, the Sr-group was separated into (1) Ba(II), Ru(III), Na(I), K(I), Fe(III), and Mo(VI) (non-sorption group); (2) Sr(II); (3) Pd(II); and (4) Zr(IV) by eluting with 2.0 M HNO 3, 0.01 M HNO 3, 0.05 M DTPA–pH 2.5, and 0.5 M H 2C 2O 4, respectively. Sr(II) adsorbed towards TODGA/SiO 2-P flowed into the second group and showed the excellent separation efficiency from others. Based on the elution behavior of the tested elements, an advanced PREC (Partitioning and Recovery of two heat generators from an acidic HLW (high activity liquid waste) by Extraction Chromatography) process was proposed.

Group partitioning of minor actinides and rare earths from highly active liquid waste by extraction chromatography utilizing two macroporous silica-based impregnated polymeric composites

Through an advanced vacuum sucking technique, two macroporous silica-based N, N, N′, N′-tetraoctyl-3-oxapentane-1,5-diamide (TODGA) and octyl(phenyl)- N, N-diisobutylcarbamoyl-methylphoshine oxide (CMPO) polymeric composites, TODGA/SiO 2-P and CMPO/SiO 2-P, were synthesized by impregnating and immobilizing the chelating agents into the pores of the SiO 2-P particles. To partition long-lived minor actinides (MA(III)), such as Am(III) and Cm(III) from highly active liquid waste (HLW), the loading and elution of 15 typically simulated fission and non-fission products from a 3.0 M HNO 3 were performed. In the first column packed with TODGA/SiO 2-P, the simulated elements were effectively separated into following groups: (1) non-adsorption group, (2) MA– lRE–FPs group, (3) hRE group, and (4) micro quantity of Zr(IV). MA(III) was predicted to flow into the second group along with Nd(III) due to their close adsorption and elution onto TODGA/SiO 2-P. In the second column packed with CMPO/SiO 2-P, after conditioning the acidity in MA– lRE–FPs containing 0.05 M DTPA to 3.0 M HNO 3, it was separated into (1) Sr(II) and Pd(II), (2) MA– mRE, (3) lRE, and (4) Zr(IV). MA(III) was believed to flow into MA– mRE effluent along with Gd(III) because of the similar adsorption towards CMPO/SiO 2-P. In terms of the position of MA(III) appeared, an improved MAREC process for long-lived MA(III) partitioning from HLW was proposed.

Palladium Removal from the Simulated Nuclear Spent Fuel Solution Using a Silica‐Based SiPyR‐N3 Anion Exchanger

The ERIX process consists of electroreduction and anion exchange for reprocessing spent nuclear fuels, in which Pd removal from spent fuel solution is necessary in the first step, since Pd can deposit at the electrode in the metallic state and interfere with the electroreduction. In the present work, the possibility in selective removal of Pd using a porous silica‐supported anion exchanger, SiPyR‐N3, containing a pyridine group has been studied. Adsorption behavior of Pd(II) from a simulated solution using a SiPyR‐N3 packed column has been investigated. The elution of Pd(II) using 0.05 M DTPA and 0.2 M thiourea was examined and compared. The SiPyR‐N3 resin showed a significantly strong adsorption for Pd(II) from HNO3 solution. The unique adsorption was considered to result from the complex formation between Pd(II) and the resin which contains the soft‐atom N with a lone pair of electrons. U(VI) and some typical simulated fission products, La(III), Pr(III), and Sr(II), showed almost no or very weak adverse impact on the Pd(II) separation, except for Re(VII) and Ru(III). The results of column experiments demonstrated that the Pd can be selectively removed from the nuclear spent fuel solution, though further work is needed to improve Pd elution.

The Effects of Simulated Fission Products in the Reduction of Hydrogen Peroxide on Simulated Nuclear Fuel Electrodes

The electrochemical reduction of hydrogen peroxide has been studied on a series of simulated nuclear fuel (SIMFUEL) materials with different amounts of simulated fission products. Steady-state polarization measurements under diffusion-compensated conditions show that the reaction proceeds with a weak dependence on applied potential and fractional reaction orders with respect to . A mechanism for reduction on SIMFUEL materials is proposed to explain these features. It is found that the activity of the material for the reduction reaction increases with the doping level. This increase can be attributed to noble metal particles present in the SIMFUEL. This enhanced reactivity is most apparent at low overpotentials, where reduction proceeds significantly faster on the noble metal particles than on the lattice.

Proficiency test for gamma spectroscopic analysis with a simulated fission product reference spectrum

Within the proficiency test programme for the radionuclide laboratories supporting the verification of the Comprehensive Nuclear-Test-Ban Treaty, a simulated gamma spectrum with the characteristics of an atmospheric nuclear test was used as reference material. The spectrum was produced by the MCNP-based Virtual Gamma Spectroscopy Laboratory (VGSL), using analysis results of a historical measurement of nuclear weapons debris as input. The method was found suitable for a proficiency test assessing laboratories’ gamma spectroscopic analysis.

Anodic Dissolution of UO2 Pellet Containing Simulated Fission Products in Ammonium Carbonate Solution

We have proposed a reprocessing process based on the dissolution of spent fuels in aqueous carbonate solution. In order to develop such a reprocessing process, anodic dissolution experiments were carried out by using a simulated spent fuel pellet in (NH4)2CO3 solution. The dissolution rate constant of the simulated fuel pellet at 323 K was estimated as 3.1 × 10-6 mol·cm−2·min−1, which is comparable to that in 5 mol·dm−3 HNO3 solution containing 0.01 mol·dm−3 HNO2 at 323 K. The dissolution rate and the current efficiency for the dissolution of simulated fuel pellet in (NH4)2CO3 solution were found to decrease with the elapse of time. Furthermore, in order to examine dissolution behavior, the surface conditions of the pellet before and after the dissolution experiments were analyzed with the scanning electron microscopy and the energy dispersive X-ray spectroscopy. It was found that neat UO2 matrix exists on the dissolved surface and a decrease in the dissolution rate and the current efficiency is caused by the increase in such surface area on the pellet. During the dissolution experiments, precipitates of the simulated fission products were observed on the pellet and in (NH4)2CO3 solution used as the electrolytic solution. Analyses of the electrolytic solution revealed that most of the simulated fission products, i.e. alkaline earth and rare earth elements, are precipitated in high ratios. From these results, it is expected that the anodic dissolution of spent fuels and the separation of fission products by precipitation can be performed simultaneously.

Removal of Pd(II), Zr(IV), Sr(II), Fe(III), and Mo(VI) from simulated high level liquid waste by extraction chromatography utilizing the macroporous silica-based polymeric materials

To eliminate a few specific elements from highly active liquid waste (HLW), two macroporous silica-based N, N, N′, N′-tetraoctyl-3-oxapentane-1,5-diamide (TODGA) and octyl(phenyl)- N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) impregnated polymeric composites, TODGA/SiO 2-P and CMPO/SiO 2-P, were synthesized. It was done by an advanced vacuum sucking technique through impregnation of the chelating agents into the pores of SiO 2-P particles support. The partitioning of Sr(II), Pd(II), Zr(IV), Mo(VI), and Cs(I) from a 3.0 M HNO 3 solution containing some typically simulated fission and non-fission products Ru(III), Rb(I), Fe(III), Na(I), K(I), and Ba(II) was performed at 298 and 323 K, respectively. In the first column packed with TODGA/SiO 2-P, Sr(II), Pd(II), and Zr(IV) were adsorbed and eluted effectively by 0.01 M HNO 3, 0.05 M DTPA-pH 2.5, and 0.5 M oxalic acid, respectively, while others showed very weak or almost no adsorption and flowed into effluent along with 3.0 M HNO 3. Cs(I), Fe(III), and Mo(VI) containing effluent was then applied to the second column packed with CMPO/SiO 2-P. Fe(III) and Mo(VI) were found to be adsorbed together and eluted utilizing 0.05 M DTPA-pH 2 at 323 K. The effective partitioning of Fe(III) was of great benefit to the evaporation, concentration, and finally geological disposal of HLW. The separation of Sr(II) and Mo(VI) was useful for the vitrification of HLW due to the Sr heat generator property and low solubility of Mo oxides in molten glass materials. Cs(I) can be separated by a novel silica-based 1,3-[(2,4-diethyl-heptylethoxy)oxy]-2,4-crown-6-calix[4]arene (Calix[4]arene-R14) impregnated polymeric composite (Calix[4]arene-R14/SiO 2-P). In terms of the results obtained, a highly specific partitioning process of Fe(III) and some specific fission products Sr(II), Pd(II), Zr(IV), Cs(I), and Mo(VI) from a simulated HLW by extraction chromatography was proposed.

Anodic Dissolution of UO2 Pellet Containing Simulated Fission Products in Ammonium Carbonate Solution

Anodic Dissolution of UO2 Pellet Containing Simulated Fission Products in Ammonium Carbonate Solution

The effects of the physical states of a simulated fission product on the linear thermal expansion of (U 0.924Ce 0.076)O 2

The linear thermal expansions of an (U 0.924Ce 0.076)O 2 pellet, doped a simulated fission product (Nd 2O 3 or Ru), were measured from room temperature to 1673 K in a flowing argon atmosphere using TMA. Nd 2O 3 and Ru represent the physical states of a fission product, a dissolved oxide and a metallic precipitate, respectively. Using the measured data, the mean coefficients of a linear thermal expansion was obtained as a function of the temperature, and the effects of the physical states of a simulated fission product on the thermal expansion were investigated. In the case of the Nd 2O 3 forming a dissolved oxide, the thermal expansion of the sample increased and the increment was proportional to the Nd contents, because the melting point of the Nd 2O 3 was lower than that of UO 2 and although the metallic precipitate hardly affected the crystal structure, the linear thermal expansion also increased with an increasing Ru contents.

Using selective precipitant for uranyl ions — Fundamental studies for evaluating the precipitant performance

We have developed a simple reprocessing process for spent FBR fuels using N-cyclohexyl-2-pyrrolidone (NCP) which has selective precipitation ability for UO 2 2+ ions. It was confirmed that NCP has sufficient precipitation ability for UO 2 2+ ions, decontamination capability (separation of UO 2 2+ from simulated fission products), and resistance to γ-ray radiation in nitric acid solutions. These findings indicate that NCP is applicable to our reprocessing process. We have also evaluated performances of other precipitants such as N-n-propyl-2-pyrrolidone (NProP), N-n-butyl-2-pyrrolidone (NBP), and N-n-butyl-2-pyridone (NBPyr). It was found that higher decontamination factors (DFs) are obtained by using NProP and NBP. This can be interpreted that the hydrophobicity of NProP and NBP is lower than that of NCP. Furthermore, we have obtained an experimental result that the resistance of NBPyr to γ-ray radiation is superior to that of NCP.

Synthesis of two novel macroporous silica-based impregnated polymeric composites and their application in highly active liquid waste partitioning by extraction chromatography

Two kinds of novel macroporous silica-based chelating polymeric adsorption materials, TODGA/SiO2-P and CMPO/SiO2-P, were synthesized by impregnating and immobilizing two chelating agents, N,N,N',N'-tetraoctyl-3-oxapentane-1,5-diamide (TODGA) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphoshine oxide (CMPO), into the pores of SiO2-P particles. To separate minor actinides (MA(III)) such as Am(III) and Cm(III), the adsorption and elution of 13 typically simulated fission products from a 3 M HNO3 were performed. It was found that in the first column packed with TODGA/SiO2-P, all of the simulated elements were separated effectively into four groups: (1) Cs(I), Mo(VI), and the most portion of Ru(III) (non-adsorption group), (2) Sr(II), small portion of Gd(III) and all of light REs(III) (MA-lRE-Sr group), (3) most of Gd(III) and all heavy RE(III) (hRE group), and (4) Zr(IV), Pd(II), and a little of Ru(III) (Zr-Pd group) by eluting with 3.0 M HNO3, 1.0M HNO3, distilled water, and 0.5 M H2C2O4, respectively, at 298 K. MA(III) was predicted to flow into the second group along with Nd(III) because of their close adsorption-elution onto TODGA/SiO2-P. In the second column packed with CMPO/SiO2-P, MA-lRE-Sr group was separated into (1) Sr(II), (2) middle RE(III) such as Gd(III), Eu(III), Sm(III), and quite small portion of Nd(III) (MA-mRE), and (3) light RE(III) such as La(III), Ce(III), and most of Nd(III) by eluting with 3.0 M HNO3 and 0.05 M DTPA-pH 2.0, respectively, at 323 K. MA(III) was believed to flow into MA-mRE group along with Gd(III) due to their similar adsorption properties towards CMPO/SiO2-P. Based on positions of MA(III) appeared in light and heavy RE(III), an improved MAREC process for MA(III) partitioning from HLW was proposed.