• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.875-883
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• 2015

The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing $^{99}Mo$. In this method, the medical isotope production system itself is used to extract $^{99}Mo$ or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of $^{99}Mo$ by irradiating targets. In this study, the neutronic performance and $^{99}Mo$, $^{89}Sr$, and $^{131}I$ production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ~1,500 Ci/wk (~325 6-day Ci) of $^{99}Mo$ at the end of a cycle.

• The Journal of Engineering Geology
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• v.23 no.4
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• pp.477-491
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• 2013

Natural radionuclides in groundwater in the Danyang area were investigated to characterize the behaviors of uranium and radon with respect to lithology and physico-chemical components, which can aid our understanding of their occurrence, properties, and origins. To this end, a total of 100 groundwater samples were collected and analyzed, and radionuclide levels were used to construct detailed concentration maps. The water type of the groundwater, assessed using a Piper diagram, is mainly Ca-Na-$HCO_3$. The concentrations of uranium range from 0.02 to $251.0{\mu}g/L$ (average, $3.85{\mu}g/L$) and only 1% exceed USEPA's MCL (Maximum Contaminant Level). Uranium is enriched in groundwaters of Cretaceous granites and Precambrian metamorphic rocks, whereas it is depleted in groundwaters of sedimentary rocks. The concentrations of radon range from 13 to 28,470 pCi/L (average, 2397 pCi/L). Only 15% of the samples exceed AMCL (Alternative Maximum Contaminant Level) of 4000 pCi/L. The radon concentration is highest in groundwater of Cretaceous granites and lowest in groundwater of sedimentary rocks. In conclusion, the distribution and occurrence of radionuclides are intimately related to the basic geological characteristics of the rocks in which the radiogenic minerals are primarily contained. The behavior of uranium is only weakly related to that of radon (correlation coefficient = 0.15). There are also weak correlations between radionuclides and the main chemical components, pH, EC, Eh, and well depth. Of note, the correlation coefficient between radon and $SiO_2$ is 0.68, and that between radon and $HCO_3$ is -0.48. Factor analysis shows that radionuclides behave somewhat independently of each other because there are no significant factors that control the behavior of chemical components as well as radionuclides. The detailed concentration maps during this study will be used to establish useful database of radionuclide distribution and geological properties throughout Korea.

• Journal of the Korean Society of Radiology
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• v.15 no.1
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• pp.37-43
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• 2021

Republic of Korea has many nuclear facilities in the country, and Democratic People's Republic of Korea(North Korea) locates in the surrounding country. Therefore, it is necessary to construct the target facility's nuclear forensic data in a preemptive response to the changing international situation. For this reason, this study suggests "signature" materials used to understand the origins and sources of nuclear and other radioactive materials, taking into account domestic nuclear facilities and the nuclear fuel cycle. In domestic, pressurized light water reactors and pressurized heavy water reactors are in operation, and enriched and natural uranium are used as fuels. In the front-end fuel cycle, the signature materials can be nature uranium and UF6 in the uranium enrichment process. The domestic back-end fuel cycle adopts a non-circulating cycle excluding the reprocessing process, and the primary signature material is spent nuclear fuel. According to IAEA recommendation, the importance of these materials as the signature and characteristic contents are suggested in this study. To prove the integrity of nuclear material and build a national nuclear forensics library, it is necessary to grasp the signature material and acquire the characteristic data considering the domestic nuclear facilities and the nuclear fuel cycle.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3913-3923
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• 2023

Todays, medium energy resolution detectors are preferably used in radioisotope identification devices(RID) in nuclear and radioactive material categorization. However, there is still a need to develop or enhance « automated identifiers » for the useful RID algorithms. To decide whether any material is SNM or NORM, a key parameter is the better energy resolution of the detector. Although masking, shielding and gain shift/stabilization and other affecting parameters on site are also important for successful operations, the suitability of the RID algorithm is also a critical point to enhance the identification reliability while extracting the features from the spectral analysis. In this study, a RID algorithm based on Bayesian statistical method has been modified for medium energy resolution detectors and applied to the uranium gamma-ray spectra taken by a LaBr3:Ce detector. The present Bayesian RID algorithm covers up to 2000 keV energy range. It uses the peak centroids, the peak areas from the measured gamma-ray spectra. The extraction features are derived from the peak-based Bayesian classifiers to estimate a posterior probability for each isotope in the ANSI library. The program operations were tested under a MATLAB platform. The present peak based Bayesian RID algorithm was validated by using single isotopes(²⁴¹Am, ⁵⁷Co, ¹³⁷Cs, ⁵⁴Mn, ⁶⁰Co), and then applied to five standard nuclear materials(0.32-4.51% at.²³⁵U), as well as natural U- and Th-ores. The ID performance of the RID algorithm was quantified in terms of F-score for each isotope. The posterior probability is calculated to be 54.5-74.4% for ²³⁸U and 4.7-10.5% for ²³⁵U in EC-NRM171 uranium materials. For the case of the more complex gamma-ray spectra from CRMs, the total scoring (ST) method was preferred for its ID performance evaluation. It was shown that the present peak based Bayesian RID algorithm can be applied to identify ²³⁵U and ²³⁸U isotopes in LEU or natural U-Th samples if a medium energy resolution detector is was in the measurements.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3464-3466
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• 2007

Power uprate is the process of increasing the maximum power level at which a commercial nuclear power plant may operate. Power uprate applications(113 units) for NPPs(Nuclear Power Plants) were recently approved in the United States. Utilities have been using power uprates since the 1970s as a way of increasing the power output of their nuclear plants. To increase the power output of a reactor, typically more highly enriched uranium fuel and/or more fresh fuel is used. This enables the reactor to produce more thermal energy and therefore more steam, driving a turbine generator to produce electricity. In this paper, the propriety of power uprate is explained through the review on the power uprate method and the changes of the physical parameters due to power uprate. The analysis results showed that the CDF(Core Damage Frequency) and LERF(Large Early Release Frequency) are affected in the current probabilistic safety assessment (PSA) model.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.125-146
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• 2014

This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.351-359
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• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• Journal of Energy Engineering
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• v.13 no.4
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• pp.242-258
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• 2004

The duplex burnable poison absorbers concept was suggested by Korea Atomic Energy Research Institute. This BP rod is composed of inner region of natural U-Gd$_2$O$_3$ and outer shell of enriched UO$_2$-Er$_2$O$_3$. It is expected that this burnable absorber has same reactivity control capability with gadolinia burnable absorber used in extened fuel cycle. In order to evaluate the nuclear feasibility of duplex BPs, the nuclear design characteristics were compared with that of four types of burnable absorbers; gadolinia, erbia, IFBA, dysprosia duplex BP on 24 months fuel cycle for Korean Standard Nuclear Power plants. According to the evaluation results of nuclear characteristics, the duplex BPs were better than other BPs on k-infinitives, reactivity holddown worth (RHW), pin power peaking and moderator temperature coefficient (MTC). The possibility of nuclear core design was also confirmed based on the optimized fuel assemblies which were searched for a sensitivity analysis. Characteristics of core design with duplex BPs was compared with that of reference core with gadolinia BPs for cycle length, power peaking and MTC. The duplex BP core had a little longer cycle length by 4 to 7 days because of increased amount of fissile in enriched uranium at the outer shell of duplex BP In case of power peaking F$\_$Q/ of duplex BP core was reduced from 1.5773 to 1.5335. MTC was also less -0.48 pcm/C than that of reference core. Finally, evaluation of fuel cycle economy was performed for the manufacturing feasibility test and fuel cost evaluation with duplex BPs. Fuel cycle economy of duplex BP core almost was equivalent with that of gadolinia BP core.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

• The Journal of the Petrological Society of Korea
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• v.17 no.3
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• pp.179-190
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• 2008

The distribution characteristics of some major environmental radionuclides ($^{40}K$, $^{228}Ac$, $^{226}Ra$, $^{222}Rn$) and U in rocks, soils and soil gas were studied at Geumjeong-Gu, Busan, Korea. The activities of radionuclides in granitic rocks are decreased in the odor of $^{40}K$>thorium decay series>uranium decay series. This reveals that Th was relatively more enriched in granitic rocks than U. The U content and activity of $^{226}Ra$ and $^{228}Ac$, however, don't reflect the fractionation sequence of granitic rocks in the study area. The activities of all these radionuclides and U content in soils are generally higher than in rocks, and their distribution in rocks, soils and soil gas show very low co-relationship. These facts indicate that the activities of radionuclides in soil and soil gas were greatly affected by leaching and adsorption properties of the radionuclides and their parents during weathering and pedogenetic process rather than their concentrations in parent rocks.