Field test around Fukushima Daiichi nuclear power plant site using improved Ce:Gd3(Al,Ga)5O12 scintillator Compton camera mounted on an unmanned helicopter

An improved light-weight Compton camera exhibiting low power consumption was developed to be mountable on an unmanned helicopter to detect cesium radiation hot spots and confirm the decontamination effect of cesium-affected areas. An increase in the Ce:Gd₃(Al,Ga)₅O₁₂ scintillator array from 4 × 4 to 8 × 8 and expansion of the interlayer distance enhanced the detection efficiency and angular resolution, respectively. Measurements were performed over the Ukedo riverbed in Namie, Fukushima Prefecture (Japan). The helicopter's flight path and speed were pre-programmed to lines interspaced by 5 and 10 m intervals and 1 m/s, respectively, facilitating measurements over areas of 65 × 60 m² and 65 × 180 m² at a height of 10 m for approximately 20 and 30 min, respectively. Results provided accurate ambient dose equivalent rate maps at a height of 1 m with an angular resolution corresponding to a position resolution of approximately 10 m from a height of 10 m. Hovering flights were executed over hot-spot areas for 10–20 min at a height of 5–20 m. Gamma-ray images of these hot spots were obtained using a reconstruction software. Comparison between position-shifted measurement results showed that the angular resolution coincided with that evaluated in the laboratory (approximately 10°).     

Recovery of neutron-irradiation-induced defects of Al2O3, Y2O3, and yttrium-aluminum garnet

SiC fiber-reinforced SiC matrix composites (SiC_f/SiC) are considered as one of the candidates for blanket materials in future fusion reactors and as an advanced fuel cladding material for next-generation fission reactors. Generally, the densification of SiC needs sintering additives and oxides such as Al₂O₃, Y₂O₃, and yttrium-aluminum garnet (YAG, Y₃Al₅O₁₂), which are frequently added to SiC. However, the effects of neutron irradiation on sintering additives are still unclear. In this study, we performed the neutron irradiation of Al₂O₃, Y₂O₃, and YAG at fluences up to 2.0–2.5 × 10²⁴ n/m² (E > 0.1 MeV) at 60–90 °C. The isochronal recovery of the macroscopic volume of Al₂O₃ against annealing temperature showed smooth and continuous shrinkage at a temperature of up to 1200 °C, and the volume slightly increased above that temperature. In contrast, the volume of Y₂O₃ showed quick shrinkage at the low temperature range, and slower and smooth recovery was observed up to ～1100 °C. In the case of YAG, the recovery of volume occurred in a step-wise manner at 600–750 °C, and continuous shrinkage occurred at temperatures lower and higher than that temperature range. The activation energies for the macroscopic volume recoveries of three oxides were obtained from the Arrhenius plots of the rate coefficients. Two-stage recovery was observed for Al₂O₃, whereas more complicated recovery processes were suggested for Y₂O₃ and YAG.     

Investigation of kinetic recovery process in low dose neutron-irradiated nuclear graphite by thermal annealing

To investigate the kinetic recovery process of low dose neutron-irradiated graphite, nuclear-grade isotropic graphite IG-110U and ETP-10 were neutron irradiated using the JMTR up to 1.38 × 10²³ n/m² (E_n > 1 MeV) at ~473 K. In-situ measurement of macroscopic length was conducted during the isothermal and isochronal annealing process from room temperature up to 1673 K. From room temperature to 773 K for IG-110U, and to 1023 K for ETP-10, macroscopic lengths, lattice parameters, and unit cell volumes of both specimens recovered to their pre-irradiation values, and this recovery process subdivided into two stages. The activation energies of macroscopic volume recovery at 523–673 K and 673–773 K were determined to be ~0.22 eV and ~0.57 eV for IG-110U, respectively; ~0.13 eV and ~2.59 eV at 523–923 K and 923–1023 K for ETP-10, respectively. The migration of not only single interstitials but also interstitials dissociated from submicroscopic interstitial groups along basal planes followed by vacancy-interstitial recombination play a dominant role in the first stage. The second stage is suggested to proceed via the motion of carbon groups along basal planes for IG-110U, and migration of single interstitials along the c-axis for ETP-10. During 773 K or 1023 K up to 1673 K, macroscopic length continuously shrank with decreasing shrinking rate, even with a turnaround to swell at 1173 K for IG-110U.     

Oxidation of Uranium Dioxide

The oxidation of UO₂ was studied by thermogravimetry and X-ray diffraction. It was clarified that the thermal history covering the first stage of the oxidation from UO₂ to U₃O UO₇ significantly influenced the rate of the oxidation of the second stage from U₃O₇ to U₃ O₈.

The entire oxidation reaction proceeded in what to all appearances, was a single stage when the specimen temperature was raised rapidly, whereas at slower rates of heating up, two distinct stages of oxidation were observed, separated by an intermediate induction period. These findings suggest the existence of a close connection between the rate of formation of the U₃O₇ phase and the rate of the subsequent oxidation of this phase: A slow formation of U₃O₇ would tend to prolong the induction period preceding the second stage of the oxidation. A similar effect was observed also with annealing of the intermediate U₃O₇ at 200°C: The increase of annealing time prolonged the induction stage.

The rate of the second stage oxidation was fairly well expressed by Johnson & Mehl's equation, log (1/(1-y/)=(1/2.303)kⁿtⁿ . The time exponent n in this equation varied in the range of 1.0~2.5, and the rate constant k of 1.15×10^?4~2.04 ×10^?1 min^?1, depending on the experimental conditions.     

Application of N,N-di(2-ethylhexyl)butanamide for mutual separation of U(VI) and Pu(IV) by continuous counter-current extraction with mixer-settler extractors

Continuous counter-current extraction using N,N-di(2-ethylhexyl)butanamide (DEHBA) as an extractant was performed with mixer-settler type extractors consisting of U–Pu extraction, scrub, U recovery, Pu back-extraction, and U back-extraction steps. The feed solution used in the continuous counter-current extraction was 3 mol/dm³ (M) nitric acid containing U, Pu, and simulated fission products of Sr, Ba, Zr, Mo, Ru, Rh, Pd, and Nd. More than 99.9% of U and Pu in the feed was extracted by 1.9 M DEHBA at the U–Pu extraction step with negligible extraction of Sr, Ba, Mo, Ru, Rh, and Nd. The extracted Pu was back-extracted via contact with 0.3 M nitric acid in the Pu back-extraction step, and the ratio of Pu distributed to the Pu fraction stream was ～ 82%. It was confirmed that 1.9 M DEHBA effectively recovered U in the U recovery step, and the ratio of U in the Pu fraction stream was less than 1%. The extracted U was back-extracted in the U back-extraction step, and more than 98% of U was recovered in the U fraction stream.     

Calcium isotope separation by band chromatography using 18-crown-6-ether resin

To develop the ⁴⁸Ca enrichment process, a feasibility study on a band chromatography was made using 9 M HCl solution and crown ether resin synthesized in porous silica beads. Prior to the chromatographic experiments, distribution coefficients, Kd, of Ca²⁺ and Sr²⁺ were measured at different concentrations of these ionic species. The frontal boundary of the chromatography was made by a usual manner of the breakthrough mode of calcium feeding, and the rear boundary was made by introducing strontium as a following ion on the basis of the observed Kd values. It was confirmed that the heavy isotope ⁴⁸Ca was depleted in the rear boundary region, while ⁴⁸Ca was enriched in the front boundary region. The values of separation coefficient ε (= α – 1) in three chromatographic operations at different temperatures were observed as 2 × 10^?3 ~ 3 × 10^?3. The separation coefficients observed in the front boundary regions, where ⁴⁸Ca was enriched, agreed with those observed in the rear boundary regions, where ⁴⁰Ca was enriched.     

Helium bubbles and trace of lithium in B4C control rod pellets used in JOYO experimental fast reactor

B₄C pellets used in the control rod of the experimental fast reactor ‘JOYO’ with different ¹⁰B burnups from lower than 10 × 10²⁰ captures/cm³ to 80 × 10²⁰ captures/cm³ and irradiated at less than 800 °C were examined by transmission electron microscopy (TEM). In a B₄C pellet irradiated in an irradiation capsule of ‘JOYO’ at 800 °C up to 30 × 10²⁰ captures/cm³, intragranular helium bubbles appeared in flat plate-shapes with the plane of the plate parallel to the (111) rhombohedral plane. However, in the other specimens that were taken from an actual control rod, the helium gas formed very tiny spherical intergranular bubbles with a diameter of a few nanometers . These tiny bubbles make wavy arrays roughly parallel to the (111) plane. The B₄C specimens were heated on a TEM in situ heating holder up to 1040 °C for 10 min. Clustering of tiny bubbles was observed, but did not extend to the plate-shaped bubbles. In high burnup specimens, large bubbles/cracks were rarely found along the {100} planes, which may correspond to the amorphous bands caused by the slip. While heating the specimens in TEM over 800 °C, liquid phases of lithium-bearing compoundsappeared on the surface of specimen.     

Neutron capture cross section measurements of 151,153Eu using a pair of C6D6 detectors

We have measured the neutron capture cross sections of ¹⁵¹Eu and ¹⁵³Eu by the time-of-flight (TOF) method in the range from 0.005 eV to keV region using the Kyoto University Research Reactor Institute - Linear Accelerator (KURRI-LINAC). We employed a pair of C₆D₆ liquid scintillators for the prompt capture γ-ray measurement. The pulse-height weighting technique was employed to obtain the capture yields from the γ-ray spectra of ^151,153Eu. The obtained thermal cross sections at 0.0253 eV are 9051 ± 683 b for ¹⁵¹Eu and 364 ± 44 b for ¹⁵³Eu, respectively. The resonance integrals have been derived as 3490 ± 162 b for ¹⁵¹Eu and 1538 ± 106 b for ¹⁵³Eu.

The obtained capture cross sections were compared with the previously reported experimental data and the evaluated data. The evaluated data in JENDL-4.0 and JEFF-3.2 show good agreement with the present experiment results of ¹⁵¹Eu, however, the evaluated data in ENDF/B-VII.1 are larger than the present experiment results of ¹⁵¹Eu about 10% to 20% in the energy region from 0.03 to 0.2 eV. For the neutron capture cross sections of ¹⁵³Eu, the evaluated data in ENDF/B-VII.1 and Widder's data are in good agreement with the present results in the energy region below 0.35 eV.     

Equilibrium and kinetic studies of selective adsorption and separation for strontium using DtBuCH18C6 loaded resin

A crown ether loaded resin was prepared by successive impregnation and fixing the 4′,4′(5″)-di(tert-butylcyclohexano)-18-crown-6 (DtBuCH18C6) and its molecule modifier, 1-dodecanol, onto the porous silica/polymer composite support (SiO₂-P particles). The characterization of DtBuCH18C6 loaded resin was examined by thermal gravimetry and differential thermal analysis and electron probe microanalysis. The adsorption behavior of Sr(II), Cs(I), Ru(III), Pd(II), La(III), Nd(III), Sm(III), Gd(III), Zr(IV), and Mo(VI) was investigated by the batch method. Furthermore, the column test for Sr (II) was performed. The batch experiments were carried out by varying the shaking times, HNO₃ concentration, and initial concentration of metal ions. A relatively large K _d value above 182 cm³/g for Sr(II) was obtained in the presence of 3 M HNO₃. In contrast, the K _d values of Cs(I), Ru(III), Pd(II), La(III), Nd(III), Sm(III), Gd(III), Zr(IV), and Mo(VI) were considerably lower than 10 cm³/g. The adsorption of Sr(II) was found to be controlled by chemisorption mechanism, and followed a Langmuir-type adsorption equation. The breakthrough curve of Sr(II) had S-shaped profile, and the elution percentage was estimated to be 99.9% by using the eluent of H₂O.     

Thermodynamic model for Zr solubility in the presence of gluconic acid and isosaccharinic acid

Zr solubility in the presence of gluconic acid (GLU) and isosaccharinic acid (ISA) was investigated as a function of hydrogen ion concentration (pH_c) and the total concentration of GLU or ISA. The dependence of the increase in Zr solubility on the pH_c and GLU concentration suggested the existence of Zr(OH)₄(GLU)₂^2? in the neutral pH region and Zr(OH)₄(GLU)(GLU_-H)^3? in the alkaline pH region above pH_c 10 as the dominant species in the presence of 10^?3–10^?1 mol/dm³ (M) GLU. In the presence of ISA, the dominant species Zr(OH)₄(ISA)₂^2? and Zr(OH)₄(ISA)(ISA_-H)^3? were proposed to occur in the neutral and alkaline pH regions, similar to those found in the presence of GLU. From X-ray diffraction analysis, the solubility-limiting solid phase in the presence of GLU and ISA was considered to be Zr(OH)₄(am). The formation constants of the Zr gluconate and isosaccharinate complexes were determined by least-squares fitting analysis of the solubility data, and the obtained values were discussed in comparison with those of tetravalent actinides.     

Isotope dilution inductively coupled plasma mass spectrometry for determination of 126Sn content in spent nuclear fuel sample

The ¹²⁶Sn content in a spent nuclear fuel solution was determined by isotope dilution inductively coupled plasma mass spectrometry (ID-ICP-MS) for its inventory estimation in high-level radioactive waste. A well-characterized irradiated UO₂ fuel sample dissolved in a hot cell was used as a sample to evaluate the reliability of the methodology. Prior to the ICP-MS measurement, Sn was separated from Te (¹²⁶Te), which causes major isobaric interference in the determination of ¹²⁶Sn content, along with highly radioactive coexisting elements, such as Sr (⁹⁰Sr), Y (⁹⁰Y), Cs (¹³⁷Cs) and Ba (^137m Ba), using an anion-exchange column. The absence of counts attributed to Te at m/z = 125, 128, and 130 in the Sn-containing effluent (Sn fraction) indicates that Te was completely removed from the anion-exchange column. After washing, Sn retained on the column was readily eluted with 1 M HNO₃ accompanied with approximately 80% of the Cd and 0.03% of the U in the initial sample. Owing to the presences of Cd and U in Sn fraction, the measurements of ¹¹⁶Sn and ¹¹⁹Sn were affected by the isobaric ¹¹⁶Cd and the doubly charged ²³⁸U²⁺ion, resulting in the positive bias of the determined values. With the exception of the isotopic ratios including ¹¹⁶Sn and ¹¹⁹Sn, ¹¹⁷Sn/¹²⁶Sn, ¹¹⁸Sn/¹²⁶Sn, ¹²⁰Sn/¹²⁶Sn, ¹²²Sn/¹²⁶Sn and ¹²⁴Sn/¹²⁶Sn were successfully determined and showed good agreement with those obtained through ORIGEN2 calculations. The measured concentration of ¹²⁶Sn in the spent nuclear fuel sample solution was 0.74 ± 0.14 ng/g, which corresponds to 23.0 ± 4.5 ng per gram of the irradiated UO₂ fuel (excluding the presence of ¹²⁶Sn in the insoluble residue). The results reported in this paper are the first experimental values of ¹²⁶Sn content and isotope ratios in the spent nuclear fuel solution originating in spent nuclear fuel irradiated at a nuclear power plant in Japan.     

Theoretical study of the adsorption of Cs,Cs+, I,I−, and CsI on C60 fullerene

Theoretical investigation for the adsorption of the cesium atom (Cs), the cesium iodide molecule (CsI), the iodine atom (I), the cesium cation (Cs⁺), and the iodide anion (I^?) onto the surface of a single fullerene molecule (C₆₀) are reported. A hybrid exchange–correlation functional using the Coulomb-attenuating method (CAM-B3LYP) is employed. The adsorption energies, i.e., the opposite of enthalpy change through adsorption, are calculated to be 143, 12, 9, 46, and 49 kJ mol^?1 for Cs, CsI, I, Cs⁺, and I^?, respectively. The equilibrium constant for Cs is calculated to be 7×10³ atm^?1 at the temperature of 1000 K and is seven orders of magnitude higher than that for CsI, indicating that the C₆₀ molecule adsorb the Cs atom highly selectively against the CsI molecule.     

The FeSe2(cr) solubility determined by solubility experiments of Se co-existing with Fe

To determine the equilibrium constant for ferroselite (FeSe₂(cr)) dissolution reaction, FeSe₂(cr) solubility experiments were performed at 298 ± 1 K from both the over- and under-saturation directions with Fe–Se precipitates that were aged at 348 K. X-ray diffraction (XRD) analysis detected only FeSe₂(cr) as the Se solid phase in the equilibrated precipitates. The E_h values of the equilibrated suspensions ranged from ?188.6 to ?4.9 mV vs. standard hydrogen electrode (SHE) and the pH values ranged from 6.00 to 8.76. Based on the available thermodynamic data, Se₄^2? and Fe²⁺ are thermodynamically stable within this E_h–pH range. Agreement between the solubility data obtained from the over- and under-saturation directions lends credence to the attainment of equilibrium at 298 ± 1 K. The thermodynamic interpretations using the specific ion interaction theory (SIT) model showed that E_h values and the concentrations of Se and Fe are well represented by the 2FeSe₂(cr) solubility reaction (2FeSe₂(cr) ? 2Fe²⁺ + Se₄^2? + 2e^?) with log₁₀K^○ = ?17.09 ± 0.28. The obtained log₁₀K^○ value falls within the uncertainty limits of the log₁₀K^○ value calculated from the available thermodynamic data.     

Adsorption of Molybdenum Hexafluoride on Magnesium Difluoride for Uranium Purification in FLUOREX Reprocessing

The nuclear fuel reprocessing method FLUOREX is a hybrid system based on fluoride volatility and using solvent extraction. Spent nuclear fuel is fluorinated, and most of the uranium is recovered as UF₆ gas. UF₆ contains some volatile fission product (FP) fluorides, so we considered their elimination from UF₆ by adsorbing them on fluoride adsorbents. We experimentally examined the adsorption of MoF₆ on MgF₂ adsorbent; MoF₆ is present as a volatile FP fluoride in UF₆ produced by the fluorination of spent nuclear fuel. The adsorption isotherm of MoF₆ adsorption on MgF₂ was obtained at MoF₆ pressures from 10^?4 to 50 kPa. The saturated adsorption amount was 1:3 ± 0:4 mg/m² at MoF₆ pressures from 10^?4 to 1 kPa. At MoF₆ pressure of about 10^?3 kPa, the saturated adsorption amount had no dependence on adsorption temperatures from 398 to 463 K. We deduced that MoF₆ was adsorbed as a monomolecule layer on the MgF₂ surface at MoF₆ pressures from 10^?4 to 1 kPa, and the MoF₆ partial pressure in UF₆ could be decreased below 1 × 10^?4 kPa, which is the specific MoF₆ concentration for the reenrichment process.     

Determination of the Be half-life by multicollector ICP-MS and liquid scintillation counting

A new method was designed and used for determining the half-life of the isotope ¹⁰Be. The method is based on (1) accurate ¹⁰Be/⁹Be measurements of ⁹Be-spiked solutions of a ¹⁰Be-rich master solution using multicollector ICP mass spectrometry (MC-ICP-MS) and (2) liquid scintillation counting (LSC) using the CIEMAT/NIST method for determining the activity concentrations of the solutions whose ¹⁰Be concentrations were determined by mass spectrometry. Important requirements for the success of this approach (a) was the previous coating of glass ampoules filled for counting experiments with ⁹Be, thereby reducing the risk of the adsorptive loss of ¹⁰Be; (b) the removal of Boron from solutions to be measured by MC-ICP-MS by cation chromatography without the introduction of mass fractionation and (c) the accurate determination of the mass bias of ¹⁰Be/⁹Be measurements by ICP-MS which are always affected by the space-charge effect. The mass bias factor was determined to be 1.1862 ± 0.071 for ¹⁰Be/⁹Be from careful fitting and error propagation of ratios of measured Li, B, Si, Cr, Fe, Cu, Sr, Nd, Hf, Tl and U standard solutions of known composition under the same measurement conditions. Employing this factor, an absolute ¹⁰Be/⁹Be ratio of 1.464 ± 0.014 was determined for a first dilution of the ¹⁰Be-rich master solution. This solution is now available as an absolute Be ratio standard in AMS measurements. Finally, a half-life of (1.386 ± 0.016) My (standard uncertainty) was calculated. This value is much more precise than previous estimates and was derived from a fully independent set of experiments. In a parallel, fully independent study using the same master solution, Korschinek et al. [35] have determined a half-life of (1.388 ± 0.018) My. The combined half-life and uncertainty amounts to (1.387 ± 0.012) My. We suggest the use of this value in nuclear studies and in studies that make use of cosmogenic ¹⁰Be in environmental and geologic samples.     

Vertical distributions of Iodine-129 and iodide in the Chukchi Sea and Bering Sea

ABSTRACT

Currently, most Iodine-129 (¹²⁹I) present in the environment is anthropogenic; it has been released from European nuclear reprocessing plants and transported into the northern part of the Arctic Ocean via ocean circulation. This study investigates the ¹²⁹I distribution in the Chukchi Sea and Bering Sea by considering water mass structure. The range of ¹²⁹I concentration is 0.89–5.03 × 10⁷ atoms L^?1 in the studied area, which is considered to be the fallout level. ¹²⁹I was found to be distributed almost uniformly. An increase in ¹²⁹I concentration levels caused by high ¹²⁹I water inflow from the Atlantic Ocean was not observed. Additionally, we revealed the vertical distribution of iodide, one chemical form of iodine, from the Bering Shelf area to the Chukchi Sea for the first time. The range of iodide concentrations was found to be 10–157 nM. Moreover, an increasing tendency of iodide near sea bottom was observed, suggesting that iodide is released from sediments at the bottom of the sea.     

Release of iodine from sintered thoria pellets at low temperatures

Post-irradiation thermal release behaviour of ¹³¹I from low-dose irradiated sintered ThO₂ pellets (~5.5 × 10²⁰ fissions/ m³) of average density 79% and 90% TD has been studied in the temperature range 723–1373 K. The possible mechanism of release has been discussed. The kinetics of burst release have been worked out and presented. The release behaviour of ¹³¹I has also been studied from high-dose irradiated ThO₂ pellets (~1.1 × 10²³ fissions/m³). These have shown an increase in diffusivity.     

Atmospheric 85Kr and 133Xe activity concentrations at locations across Japan following the Fukushima Dai-ichi Nuclear Power Plant accident

Atmospheric ⁸⁵Kr and ¹³³Xe activity concentrations were determined from weekly air samples collected at Sapporo, Akita and Chiba, Japan, throughout 2011. The results demonstrated that the Fukushima Dai-ichi Nuclear Power Plant accident in early March 2011 resulted in high ¹³³Xe activity concentrations as well as elevated levels of ⁸⁵Kr activity; there was a striking increase in the concentrations of both isotopes over the week running from 14 to 22 March as the radioactive plume released from the plant was captured. At Chiba, following the accident, the ⁸⁵Kr activity concentration increased from 1.38 to 17.7 Bq/m³, while the ¹³³Xe levels increased from below the minimum detectable concentration (MDC ≤ 1.9 × 10^?3 Bq/m³) to 1.3 × 10³ Bq/m³. Conversely, at Sapporo and Akita, high ⁸⁵Kr activity concentrations were not observed, due to differences in air transportation mechanisms based on wind directions. Duplicate samples were collected at Chiba to allow the simultaneous analyses of ⁸⁵Kr and ¹³³Xe at the Japan Chemical Analysis Center and the Bundesamt für Strahlenschutz in Germany and the results were in good agreement. The external effective radiation doses resulting from ⁸⁵Kr and ¹³³Xe releases following the accident were estimated to be approximately 7.0 × 10^?3 and 1.3 μSv, respectively, based on the activity concentrations of these nuclides from March to June in 2011 at Chiba.     

In-Line Uranium Monitoring in Reprocessing Waste Solution by Time-Resolved Laser-Induced Fluorometry

In-line monitoring by fluorometry of uranium concentration in reprocessing waste solution has been realized. The reduction of U0₂ ²⁺ fluorescence by coexisting ions and solution temperature can be corrected by measuring fluorescence lifetime, absorbance of excitation beam wavelength and absorbance of fluorescence wavelength. The method applicability was examined by using a sample solution simulating the waste solution of the codecontamination process. When the coexisting ion concentrations were increased in the sample solution which included 50 mg//of uranium, the corrected value of the uranium concentration was constant in the range of 0~1.5 times the coexisting ion concentrations. When the temperature was changed in the range of 30~45°C, the corrected value was also constant. The precision of the correcting method was ±15%'. These results verified that the correcting method could be applied to in-line monitoring of uranium concentration even though species and amounts of the coexisting ions and solution temperature were changed in the waste solution.