• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2660-2669
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• 2022

The attractiveness for weapons usage of the proposed fuel cycle for the PT-SCWR was evaluated in this study using the Figure-of-Merit methodology. It was compared to the attractiveness of other fuel cycles namely, Low Enriched Uranium (LEU), U/Th, Re-enriched Reprocessed Uranium (RepU), and Pu/Th/U. The optimal content of natural uranium, which can be added to Pu/Th to render the produced U-233 unattractive, was found to be 9%. A ranking system to compare the attractiveness of the various fuel cycles is proposed. RepU was found to be the most proliferation resistant fuel cycle for the first 100 years,while, the least proliferation resistant fuel cycle was the originally proposed Pu/Th one. The reactivity feedback coefficients were calculated for all proposed fuel cycles. All studied reactivity coefficients have the same sign implying that all the fuel cycles will behave neutronically in a similar way. The Pu/Th/U fuel was found to have the most negative value of the Coolant Void Reactivity which will help to restore the core to a safe status faster in case of a loss-of-coolant accident. The fuel and moderator temperature coefficients did not show significant differences between the fuels studied.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.140-144
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• 2018

Isotopic analyses of plutonium and low-enriched uranium mixtures with particle sizes of $0.6-3.3{\mu}m$ were performed using thermal ionization mass spectrometry with a continuous heating method to verify its effectiveness for the accurate analysis of minor isotopes without sample pretreatment. The mixed particles used in this study were prepared from a mixed solution of plutonium (SRM 947) and uranium (U010, $^{235}U$ 1% enriched) reference materials. The isotope ratios for plutonium in the individual mixed particles, including $^{238}Pu/^{239}Pu$, $^{241}Pu/^{239}Pu$ as well as $^{240}Pu/^{239}Pu$, and $^{242}Pu/^{239}Pu$, were in good agreement with the certified values despite the isobaric interference of $^{238}U$ and $^{241}Am$. The isotope ratios for uranium in the mixed particles also agreed well with the certified values within the range of error. However, the isotope ratios for minor isotopes, such as $^{234}U$ and $^{236}U$, in the particles with diameters of less than approximately $1.8{\mu}m$ could not be measured because numbers of $^{234}U$ and $^{236}U$ atoms in analyzed particles are too low. These results indicate that thermal ionization mass spectrometry with a continuous heating method is applicable for the analysis of trace amounts of plutonium isotopes, including $^{238}Pu$ and $^{241}Pu$, despite the presence of the respective isobars $^{238}U$ and $^{241}Am$ in the microsamples.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.353-358
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• 2006

Alpha track analysis for the determination of the trace amount of alpha emitting nuclides in a very small particle was performed as an efficient and powerful technique during safeguard inspection. Metal particles with well-defined spherical shape, size and isotopic compositions as a reference material were used to correlate the number of tracks or track diameter with an isotopic composition eventually to identify the uranium enrichment in the environmental swipe samples. Slopes in the number of tracks versus the exposure time curve provide a simple insight into the uranium enrichment of an unknown particle. Low enriched uranium metal particles result in slopes still steeper than the depleted or natural uranium metal particles. In addition, a linear relationship between track diameter and particle size Is thought to be a useful first stage analytical tool as an efficient and convenient inspection guide. The significance of the simple linear model was also judged using the usual statistical tests.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1652-1657
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• 2021

Nuclear materials can be utilized not only for peaceful uses, but also for military purposes; hence, the international community has devoted itself to the control, management and safeguarding of nuclear materials. Nuclear materials are of varying degrees of usability for development of nuclear weapons. Thus, several methods for assessing the attractiveness of nuclear materials for nuclear weapons purposes have been proposed. When these methods are applied to unirradiated depleted, natural, and low-enriched uranium (DU, NU, and LEU), they are certainly classified as non-attractive nuclear materials. However, when nuclear material attractiveness is to be evaluated for potential radiological dispersal device (RDD) uses, it is required to develop a different method for the different aspects and factors. In the present study, we derived a novel method for evaluating nuclear material attractiveness for use in RDD development. To this end, the specific activity and dose coefficient were identified as the two sub-factors, and, in consideration of those, the mass causing detrimental health effects was determined to be the main factor impacting on nuclear materials attractiveness. Based on this factor, the attractiveness of unirradiated DU, NU, and LEU for RDD use was qualitatively compared with that of ¹³⁷Cs.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3585-3593
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• 2024

Radioisotopes tend to play important roles in many fields, such as industry, healthcare, agriculture, aerospace, etc. Radioisotope production is mainly through accelerators or research reactors, and high flux research reactor is one of the most effective approaches for radioisotope production. The physical basis of preparing radioisotope relies on nuclear reactions occurring in the reactor core, which includes fission, (n,γ), (n,α), and (n,p) reaction, etc. Among them, fission and (n,γ) reaction are most important in the nuclear reactor. For example, the ⁹⁹Mo could be generated by uranium fission and extracting from the fission products, or through the radiative capture reaction from enriched ⁹⁸Mo. As for the fission method, the irradiation target is gradually transitioning from high enriched uranium (HEU) target to low enriched uranium (LEU) target due to the requirement of non-proliferation. In this paper, studies on the impacts of different fission targets on radioisotope productions are conducted. Moreover, an optimized study on the radiative capture method is performed to improve the production efficiency. It is concluded that it is advantageous to use radiative capture method to generate radioisotopes in high flux reactor, which helps to improve the specific activity with environmental friendliness.

• Geosciences Journal
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• v.22 no.6
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• pp.1001-1013
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• 2018

A new type of uranium occurrence in Korea was identified in pegmatitic and hydrothermally altered granite in the Daejeon area. The U-bearing parts typically include muscovite, pink-feldspar and sericite as alteration minerals. In this study, the geochemical characteristics and alteration age of the granitic rocks were examined to provide evidence for hydrothermally-enriched uranium. The K-Ar ages of muscovite coexisting with U-bearing minerals were determined as 123 and 128 Ma. The U-bearing rocks have relatively low ($CaO+Na_2O$), high $K_2O$ contents, and high alteration index values by major element geochemistry. The trace element geochemistry shows that the uraniferous rocks have significantly low Th/U ratios and strongly differentiated features. The rare earth element patterns indicate that the uraniferous rocks have a low total REE and LREE contents with depletion of Eu. Considering the geochemical variation of the granitic rock major, trace and rare earth elements, it can be concluded that uranium enrichment in pegmatites and altered granite should be genetically related to post-magmatic hydrothermal alteration of K-metasomatism after emplacement of the two-mica granite. This is the first report for geochemical characteristics of Mesozoic granite-related U-occurrences in South Korea. This study will help further research for uranium deposits with similarities in geological setting, mineralogy and age data between South China and Korea, and can also be expected to help solve the source problems related to high uranium concentrations in some groundwater occurring in the granitic terrane.

• Journal of Soil and Groundwater Environment
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• v.15 no.6
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• pp.9-16
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• 2010

Groundwater sampling was performed at 38 wells where they are located in the areas with high uranium and radon (marked as A and B, respectively) concentrations, which were based on the previous research results. In-situ parameters (temperature, pH, EC, Eh, DO) and natural radionuclides (uranium and radon) were analyzed to figure out the characteristics of groundwater environments. In-situ data did not show any relations to natural radionuclide data, which could be caused by groundwater mixing, depths of wells, and geological settings, etc. But the highest radon well presented relatively low temperature value and the highest uranium well presented relatively low pH values The highest uranium concentration ranging $1.14{\sim}188.19{\mu}g/L$ showed in the area of A region consisted of Jurassic two-mica granite. The areas of Jurassic biotite granite and Cretaceous granite in the A region have the uranium concentrations ranging $0.10{\sim}49.78{\mu}g/L$ and $0.36{\sim}3.01{\mu}g/L$, respectively. The uranium values from between wells of community water systems (CWSs) penetrating fractured bed-rock aquifers and personal boreholes settled in shallow aquifers near the wells of CWSs show big differences. It implies that the groundwaters of the two areas have evolved from different water-rock interaction paths that may caused by various types of wells having different aquifers. High radon activities in the area of B region composed of Precambrian gneiss showed ranging from 6,770 to 64,688 pCi/L. Even though the wells are located in the same geological settings, their rodon concentration presented different according to depth and distance.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.541-548
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• 2017

Uranium metal used for fabrication of fuel for research reactors in India is generally produced by magnesio-thermic reduction of $UF_4$. Performance of magnesio-thermic reaction and recovery and quality of uranium largely depends on properties of $UF_4$. As ammonium diuranate (ADU) is first product in powder form in the process flow-sheet, properties of $UF_4$ depend on properties of ADU. ADU is generally produced from uranyl nitrate solution (UNS) for natural uranium metal production and from uranyl fluoride solution (UFS) for low enriched uranium metal production. In present paper, ADU has been produced via both the routes. Variation of uranium recovery and crystal structure and composition of ADU with progress in precipitation reaction has been studied with special attention on first appearance of the precipitate Further, ADU produced by two routes have been calcined to $UO_3$, then reduced to $UO_2$ and hydroflorinated to $UF_4$. Effect of two different process routes of ADU precipitation on the characteristics of ADU, $UO_3$, $UO_2$ and $UF_4$ were studied here.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1517-1523
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• 2020

The widely used medical isotope technetium-99 m (^99mTc) is a daughter of Molybdenum-99 (⁹⁹Mo), which is mainly produced using dedicated research reactors from the nuclear fission of uranium-235 (²³⁵U). ^99mTc has been used for several decades, which covers about 80% of the all the nuclear diagnostics procedures. Recently, the instability of the supply has become an important topic throughout the international radioisotope communities. The aging of major ⁹⁹Mo production reactors has also caused frequent shutdowns. It has triggered movements to establish new research reactors for ⁹⁹Mo production, as well as the development of various ⁹⁹Mo production technologies. In this context, a new research reactor project was launched in 2012 in Korea. At the same time, the development of fission-based ⁹⁹Mo production process was initiated by Korea Atomic Energy Research Institute (KAERI) in 2012 in order to be implemented by the new research reactor. The KAERI process is based on the caustic dissolution of plate-type LEU (low enriched uranium) dispersion targets, followed by the separation and purification using a series of columns. The development of proper waste treatment technologies for the gaseous, liquid, and solid radioactive wastes also took place. The first stage of this process development was completed in 2018. In this paper, the results of the hot test production of fission ⁹⁹Mo using HANARO, KAERI's 30 MW research reactor, was described.