• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1300-1309
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• 2023

Simulated debris was synthesized using UO₂, Zr, and stainless steel and a heat treatment method under inert or oxidizing conditions. The primary U solid phase of the debris synthesized at 1473 K under inert conditions was UO₂, whereas a (U, Zr)O₂ solid solution formed at 1873 K. Under oxidizing conditions, a mixture of U₃O₈ and (Fe, Cr)UO₄ phases formed at 1473 K, whereas a (U, Zr)O_2+x solid solution formed at 1873 K. The leaching behavior of the fission products from the simulated debris was evaluated using two methods: the irradiation method, for which fission products were produced via neutron irradiation, and the doping method, for which trace amounts of non-radioactive elements were doped into the debris. The dissolution behavior of U depended on the properties of the debris and aqueous solution for immersion. Cs, Sr, and Ba leached out regardless of the primary solid phases. The leaching of high-valence Eu and Ru ions was suppressed, possibly owing to their solid-solution reaction with or incorporation into the uranium compounds of the simulated debris.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.34-41
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• 2024

This work presents experimental data and modelling of the release of Mo from high-burnup spent nuclear fuel (63 MWd/kgU) at two different pH values, 8.4 and 13.2 in air. The release of Mo from SF to the solution is around two orders of magnitude higher at pH = 13.2 than at pH = 8.4. The high Mo release at high pH would indicate that Mo would not be congruently released with uranium and would have an important contribution to the Instant Release Fraction, with a value of 5.3%. Parallel experiments with pure non irradiated Mo(s) and XPS determinations indicated that the faster dissolution at pH = 13.2 could be the consequence of the higher releases from metallic Mo in the fuel through a surface complexation mechanism promoted by the OH^- and the oxidation of the metal to Mo(VI) via the formation of intermediate Mo(IV) and Mo(V) species.

• Journal of Radiation Industry
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• v.18 no.1
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• pp.89-93
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• 2024

Accidents at nuclear facilities and nuclear power plants led to leaks of large amounts of radioactive substances. Of the various radioactive nuclides released, ¹³⁷Cs are radioactive substances generated during the fission of uranium. Therefore, due to the high fission yield (6.09%), strong gamma rays, and a relatively long half-life (30 years), a rapid and efficient removal method and a study of adsorbents are needed. Accordingly, an adsorbent was prepared using Prussian blue (PB), a material that selectively adsorbs radioactive cesium. As a result of evaluating the adsorption performance with the prepared adsorbent, it was confirmed that 82% of the removal efficiency was obtained, and most of the cesium was rapidly adsorbed within 10 to 15 minutes. The purpose of this study was to adsorb cesium using the Prussian blue alginate bead and to compare the change in detection efficiency according to the amount of adsorbent added for quantitative evaluation. However, in this case, it is difficult to determine the detection efficiency using a standard source with the same conditions as the measurement sample, so the efficiency change of the HPGe detector according to the different heights of Prussian blue was calculated through MCNP simulation using certified standard materials (1 L, Marinelli beaker) for radioactivity measurement. It is expected to derive a relational equation that can calculate detection efficiency through an efficiency curve according to the volume of Prussian blue, quantitatively evaluate the activity at the same time as the adsorption of radioactive nuclides in actual contaminated water and use it in the field of nuclear facility operation and dismantling in the future.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3969-3980
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• 2024

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.181-190
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• 2012

Naturally occurring radioactive materials (NORM) in building materials are main sources of external radiation exposure to the general public. The objective of this study was to assess external radiation dose in Korean dwellings due to NORM in concrete walls. Reference room model for dose assessment was made by analyzing room structure and housing scale of Korean dwellings. In addition, dose assessments were made for varying room sizes. Absorbed doses to air and effective dose rates were calculated using radiation transport code MCNPX. Assuming a reference room of $3{\times}4{\times}2.8m^3$, absorbed dose rates in air were 0.80, 0.97, 0.08 nGy $h^{-1}$ per Bq $kg^{-1}$ for uranium series, thorium series, and $^{40}K$, respectively. Effective dose rates were 0.57, 0.69, 0.058 nSv $h^{-1}$ per Bq $kg^{-1}$, respectively. Radiation dose resulting from concrete of ceiling and floor increased with room area while radiation dose from concrete of walls decreased with room area. Therefore, total radiation doses were almost the same for the varying room area from 5 to $30m^2$. Effective dose in Korean dwellings was calculated based on measurement data of NORM concentration in concrete and occupancy fraction of Korean population by location. Annual effective dose was 0.59 mSv assuming that indoor occupancy fraction was 0.89 and concentrations of uranium series, thorium series and $^{40}K$ were 26, 39, 596 Bq $kg^{-1}$, respectively. Finally, annual effective dose in Korean dwellings can be calculated by the following equation: Effective dose=indoor occupancy fraction${\times}8760\;h\;y^{-1}{\times}(0.57C_U+0.69C_{Th}+0.058C_K)$.

• Journal of Radiation Protection and Research
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• v.39 no.1
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• pp.54-60
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• 2014

IAEA has employed various types of radiation detectors - HPGe, NaI, CZT - for accountancy of nuclear material. Among them, HPGe has been mainly used in verification activities required for high accuracy. Due to its essential cooling component(a liquid-nitrogen cooling or a mechanical cooling system), it is large and heavy and needs long cooling time before use. New hand-held portable HPGe has been developed to address such problems. This paper deals with results of performance evaluation test of the new hand-held portable HPGe prototype which was used during IAEA's inspection activities. Radioactive spectra obtained with the new portable HPGe showed different characteristics depending on types and enrichments of nuclear materials inspected. Also, Gamma-rays from daughter radioisotopes in the decay series of $^{235}U$ and $^{238}U$ and characteristic x-rays from uranium were able to be remarkably separated from other peaks in the spectra. A relative error of enrichment measured by the new portable HPGe was in the range of 9 to 27%. The enrichment measurement results didn't meet partially requirement of IAEA because of a small size of a radiation sensing material. This problem might be solved through a further study. This paper discusses how to determine enrichment of nuclear material as well as how to apply the new hand-held portable HPGe to safeguard inspection. There have been few papers to deal with IAEA inspection activity in Korea to verify accountancy of nuclear material in national nuclear facilities. This paper would contribute to analyzing results of safeguards inspection. Also, it is expected that things discussed about further improvement of a radiation detector would make contribution to development of a radiation detector in the related field.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.118-128
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• 1996

The lattice parameters of stoichiometric $UO_2$ and $U_{1-y}Er_{y}O_2$ in the range of y=0.01 to y =0.33 were determined with use of X-ray diffraction data. Oxygen potentials have been measured by means of a thermogravimetric method in the range of 1200~$1500^{\circ}C$ and $10^{-14}$ $\leq$ $Po_2$ $\leq$ $10^{-3}$ for pure $UO_2$ and $U_{1-y}Er_{y}O_{2{\pm}x}$ solid solutions with y=0.02, y=0.06 and y=0.20, respectively. Their oxygen partial pressures were maintained by controlling $CO_2$/CO mixture atmosphere, and the $Po_2$ values corresponding to x of $U_{1-y}Er_{y}O_{2{\pm}x}$ solid solutions were measured with an electrolyte oxygen sensor. The lattice parameter decreases linearly with an increase in the erbium content. The change of the lattice parameter can be expressed in a linear equation of y as a($\AA$) =5.4695-0.220y for 0 $\leq$y$\leq$0.33. The experimental coefficient of y -0.220 in $U_{1-y}Er_{y}O_2$ was an intermediate value between the calculated values -0.273 and -0.156 in the case of $U^{5+}$ and $U^{6+}$, respectively. The (equation omitted) has been found to undergo abrupt increase in the range of -360 to -270 kJ/mole for y=0.06 and -320 to -220 H/mole for y=0.20, respectively, in the temperature range of 1200-$1500^{\circ}C$. (equation omitted) increases with erbium content, but the effect of the dopant for x =0.01 is less significant than that for stoichiometry. The oxygen potentials for $UO_2$ and $U_{0.98}Er_{0.02}O_{2+x}$ can be approximately represented by the $U^{5+}$/$U^{4+}$ model but those for y$\geq$ 0.06 in $U_{1-y}Er_{y}O_{2{\pm}x}$ solid solutions cannot be interpreted by the mean uranium valence model.

• Economic and Environmental Geology
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• v.12 no.1
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• pp.41-49
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• 1979

A literature review is made on the physical and chemical characteristics of clay minerals in acidic solutions from the mineralogical and hydrometallurgical viewpoints. Some of the important characteristics of clays are their ability to cation exchange, swelling, and incongruent dissolution in acidic solutions. Various clay minerals can take up metallic ions from solution via cation exchange mechanism. Generally, cation exchange capacity increases in the following order : kaolinite, halloysite, illite, vermiculite, and montmorillonite. In acidic solutions, the cation uptake such as copper by clay minerals is strongly inhibited by hydrogen and aluminum ions and thus is not economically significant factor for recovery of metals such as uranium and copper. In acidic solutions, the cation uptake is substial. Swelling is minimal at lower pH, possibly due to lattice collapse. Swelling may be controllable with montmorillonite type clays by exchanging interlayer sodium with lithium and/or hydroxylated aluminum species. The effect of add on clay minerals are : 1. Division of aggregates into smaller plates with increase in surface area and porosity. 2. Clay-acid reactions occur in the following order: (i) $H^+$ replacement of interlayer cations, (ii) removal of octahedral cations, such as Al, Fe, and Mg, and (iii) removal of tetrahedral Al ions. Acid attack initiates, around the edges of the clay particles and continued inward, leaving hydrated silica gel residue around the edges. 3. Reaction rates of (ii) and (iii) are pseudo-1st order and proportional to acid concentration. Rate doubles for every temperature increment of $10^{\circ}C$. Implications in in-situ leaching of copper or uranium with acid are : 1. Over the life span of the operation for a year or more, clays attacked by acid will leave silica gel. If such gel covers the surface of valuable mineral surfaces being leached, recovery could be substantially delayed. 2. For a copper deposit containing 0.5% each of clay minerals and recoverable copper, the added cost due to clay-acid reaction is about 1.5c/lb of copper (or 0.93 lbs of $H_2SO_4/1b$ of copper). This acid consumption by clay may be a factor for economic evaluation of in-situ leaching of an oxide copper deposit.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.207-217
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• 2011

In this study the complex formation reactions between uranium(VI) and 2,6-dihydroxybenzoate (DHB) as a model ligand of humic acid were investigated by using UV-Vis spectrophotometry and time-resolved laser-induced fluorescence spectroscopy (TRLFS). The analysis of the spectrophotometric data, i.e., absorbance changes at the characteristic charge-transfer bands of the U(VI)-DHB complex, indicates that both 1:1 and 1:2 (U(VI):DHB) complexes occur as a result of dual equilibria and their distribution varies in a pH-dependent manner. The stepwise stability constants determined (log $K_1$ and log $K_2$) are $12.4{\pm}0.1$ and $11.4{\pm}0.1$. Further, the TRLFS study shows that DHB plays a role as a fluorescence quencher of U(VI) species. The presence of both a dynamic and static quenching process was identified for all U(VI) species examined, i.e., ${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$, and $(UO_2)_3{(OH)_5}^+$. The fluorescence intensity and lifetimes of each species were measured from the time-resolved spectra at various ligand concentrations, and then analyzed based on Stern-Volmer equations. The static quenching constants (log $K_s$) obtained are $4.2{\pm}0.1$, $4.3{\pm}0.1$, and $4.34{\pm}0.08$ for ${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$, and $(UO_2)_3{(OH)_5}^+$, respectively. The results of Stern-Volmer analysis suggest that both mono- and bi-dentate U(VI)-DHB complexes serve as groundstate complexes inducing static quenching.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.83-90
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• 2020

The Graphite Isotope Ratio Method (GIRM) can verify non-proliferation of nuclear weapon by estimating the total plutonium production in a graphite-moderated reactor. Using the reactor, plutonium is generated and accumulated through the ²³⁸U neutron capture reaction, and impurities in the graphite are converted to nuclides due to the nuclear reaction. Therefore, the amount of plutonium production and concentration of the impurities are correlated. However, the plutonium production cannot be predicted using only the absolute concentration of the impurities. It can only be predicted when the initial concentration of the impurities is obtained because the concentration, at a certain time, depends on it. Nevertheless, the ratios of the isotopes in an element are known regardless of the impurity of an element in the graphite moderator. Thus, the correlation between the isotope ratio and amount of plutonium produced helps predict plutonium production in a graphite-moderated reactor. Boron, Lithium, Chlorine, Titanium, and Uranium are known as indicator elements in the GIRM. To assess whether the correlation between the indicator isotope and amount of plutonium produced is independent of the initial concentration of the impurities, four different impurity compositions of graphite were used. ¹⁰B/¹¹B, ³⁶Cl/³⁵Cl, ⁴⁸Ti/⁴⁹Ti, and ²³⁵U/²³⁸U had a consistent correlation with the cumulative plutonium production, regardless of the initial impurity concentration of the graphite, because these isotopes were not generated through the nuclear reaction of other elements. On the other hand, the correlation between ⁶Li/⁷Li and plutonium production depended on the initial concentration of the impurities in graphite. Although ⁷Li can be produced through the neutron capture reaction of ⁶Li, the (n, α) reaction of ¹⁰B was the major source of ⁷Li. Therefore, the initial concentration of ¹⁰B affected the production of ⁷Li, making Li unsuitable as an indicator element for the GIRM.