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

• Korean Journal of Mineralogy and Petrology
• /
• v.36 no.4
• /
• pp.289-302
• /
• 2023

The study area was Gangnim-myeon, Hoengseong-gun, Gangwon-do, composed of the Chiaksan gneiss complex, and it was revealed that the concentrations of uranium (U) and thorium (Th) within the groundwater of the study area exceeded their water quality standards. Hence, artificial weathering experiments were conducted to elucidate mineralogically the mechanisms of their leaching using drilling cores obtained from the corresponding groundwater aquifers. First of all, the mineralogical compositions of core samples were observed, and the results indicated that the content of clinochlore, a member of the chlorite group of minerals that can form through low- and intermediate-temperature metamorphisms, was relatively higher. In addition, the Th concentration was measured ten times higher than that of U. The results of artificial weathering experiments suggested that the Th concentrations gradually increased through the dissolution of radioactive-element-bearing minerals up to the first day, and then they tended to decrease. It could be attributed to the fact that Th was leached with the dissolution of thorite, which might be a secondary mineral, and then dissolved Th was re-precipitated as the various forms of salt, such as sulfate. Even though the U content was lower than that of Th in the core samples, the U concentration was one hundred times higher than that of Th after the weathering experiments. It is likely caused by the gradual dissolution and desorption of U included in intensively weathered thorite or adsorbed as a form of UO₂²⁺ on the mineral surface. In addition, the leaching tendency of U and Th was positively correlated with the bicarbonate concentration. However, the concentrations between U and Th in groundwater exhibited a relatively lower correlation, which might result from the fact that they occurred from different sources, as aforementioned. Among various kinetic models, the parabolic diffusion and pseudo-second-order kinetic models were confirmed to best fit the dissolution kinetics of both elements. The period that would be taken for the U concentration to exceed its drinking-water standard was inferred using the regressed parameters of the best-fitted models, and the duration of 29.4 years was predicted in the neutral-pH aquifers with relatively higher concentrations of HCO₃, indicating that U could be relatively quickly leached out into groundwater.

• Parasites, Hosts and Diseases
• /
• v.36 no.3
• /
• pp.207-212
• /
• 1998

In the present study, genomic DNAs were purified from Korean isolates (KT8, KT6, KT-Kim and KT-Lee) and foreign strains (CDC85, IR78 and NYH 286) of 1Trichomonas voslnalis, and hybridized with a probe based on the repetitive sequence cloned from T. uqfinolis to observe the genetic differences. By Southern hybridization, all isolates of T. uoSinoLis except the NYH386 strain had 11 bands. Therefore all isolates examined were distinguishable into 3 groups according to their banding patterns; i) KT8, KT6 and KT-Kim isolates had 11 identical bands such as 1 kb, 1.2 kb, 1.6 kb, 1.9 kb, 2.3 kb. 27 kb, 3.2 kb, 2.4 kb, 3.8 kb, 4.9 kb and 6.0 kb, ii) The metronidazole-resistant IR78 strain had the some bands as KT-Lee isolate at bands of 1 kb, 1.2 kb, 1.6 kb. 1.8 kb, 2.1 kb, 2.5 kb, 2.7 kb, 2.9 kb, 3.4 kb, 5.0 kb and 6.0 kb, Bands of CDC85, metronidazole-resistant strain, were similar to those of IR78 and KT-Lee, except that 3.2 kb replaced 2.9 kb. iii) NYH286 particularly had 12 bands and bun patterns were similar to IR78 with a few exceptions as follows; i) 6.2 kb in place of 6.0 kb, ii) 2.0 kb and 2.2 kb instead of 2.1 kb. Through the results obtained, genetic variance of T. uoginnlis isolates was demonstrated by Southern hybridization.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1442-1450
• /
• 2017

In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the $UO_2-Na$ interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.499-507
• /
• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• 한국전산유체공학회:학술대회논문집
• /
• 2011.05a
• /
• pp.269-274
• /
• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.06a
• /
• pp.426-427
• /
• 2005

Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical process in which the spent fuel of PWR is transformed into the uranic metal ingot. Through this process, which has been developed in KAERI since 1998, the radioactivity, the radiotoxicity, the heat and the volume of the PWR spent fuel are reduced by a quarter of the original. To demonstrate a lab-scale process and extract the data for the later pilot-scale process, a demonstration facility of ACP (ACPF) is under construction and the lab-scale demonstration is slated for 2006. To establish the safeguardability of ACPF, a safeguards system including a neutron counter based on non-destructive assay, which is named as ACP Safeguards Neutron Counter (ASNC), the ACP Safeguards Surveillance System (ASSS) which consists of two neutron monitors and five IAEA cameras, and Laser Induced Breakdown System (LIBS) have been developed and are ready to be installed at ACPF. The target materials of ACP to assay with ASNC are categorized into three types among which the first is the uranic metal ingot, the second is the salt waste and the last is $UO_2$ and $U_{3}O_8$ powders, rod cuts and hulls. The Pu content of process nuclear materials can be accounted with ASNC. The ASSS is integrated in the ACP Intelligent Surveillance Software (AISS) in which the IAEA camera images and background signals at the rear doors of ACPF are displayed. The composition of special nuclear materials of ACP can be measured with LIBS which can be a supporting measurement tool for ASNC. The conceptual picture of safeguards system of ACPF is shown in Fig. 1.

• Journal of Acupuncture Research
• /
• v.21 no.6
• /
• pp.233-248
• /
• 2004

Objective : The purpose of this study is to investigate the mechanism and effect of moxibustion with monkshood cake, slice & black pepper cake. objectively, to be used as the quantitative data through the measurement of temperatqre, and to grasp the thermodynamic characteristics of moxibustion with monkshood cake, slice & black pepper cake. Methods : We have selected of the moxibustion with monkshood cake, slice & black pepper cake. indirect moxibustion. We make a comparative study of the thermodynamic characteristics of moxibustion with monkshood cake, slice & black pepper cake. We examined combustion times, temperatures, temperature gradients in each period during a combustion of moxa. Results & Conclusions : 1. We can design the moxibustion with monkshood cake that it has thermodynamic characteristics of 173sec effective combustion time, $44^{\circ}C$ maximum temperature, $0.22^{\circ}C/sec$ ascending maximum temperature, if we use 3mm thickness or 3mm and below of monkshood cake and the moxa cone is formed the conical shape that the base diameter was 8mm, the height was 10mm, the density was $600mg/cm^3$. 2. We can design the moxibustion with monkshood cake that it has thermodynamic characteristics of 205~271sec effective combustion period time, $44.6{\sim}46.1^{\circ}C$ maximum temperature, $0.18{\sim}0.24^{\circ}C/sec$ ascending maximum temperature, if we use 3mm thickness of monkshood cake and the moxa cone is formed the conical shape that the base diameter was 8mm, the height was 10mm, the density was $480{\sim}720mg/cm^3$.

• Nuclear Engineering and Technology
• /
• v.15 no.4
• /
• pp.267-279
• /
• 1983

An analytical system has been established for scrutinizing both uranium- and plutonium-fueled lattices moderated by light water. This system consists of two primary codes. One is a unit cell program called KICC, which has theoretical foundation on the models of GAM and THERMOS incorporated with appropriate approximate treatments for various phenomena, whereas the other is a multi-dimensional diffusion-depletion program entitled KIDD. The adequacy of this system is verified by performing extensive benchmark calculations on a variety of critical experiments. The average value of effective multiplication factors for the selected nineteen UO$_2$ critical experiments of heterogeneous lattice structure is calculated to be 1.0006 with a standard deviation of 0.0039. Power distributions have also been calculated for some critical experiments fueled with both uranium and plutonium of varying concentrations. The maximum percentage difference between the measured and calculated power distributions appears to be less than 5%. This result, together with the previously reported result, illustrates that the KICC/KIDD system is a very effective tool for the analysis of a light water reactor core.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3059-3072
• /
• 2022

A Jacobian-Free Newton Krylov Two-Nodal Coarse Mesh Finite Difference algorithm based on Nodal Expansion Method (NEM_TNCMFD_JFNK) is successfully developed and proposed to solve the three-dimensional (3D) and multi-group reactor physics models. In the NEM_TNCMFD_JFNK method, the efficient JFNK method with the Modified Incomplete LU (MILU) preconditioner is integrated and applied into the discrete systems of the NEM-based two-node CMFD method by constructing the residual functions of only the nodal average fluxes and the eigenvalue. All the nonlinear corrective nodal coupling coefficients are updated on the basis of two-nodal NEM formulation including the discontinuity factor in every few newton steps. All the expansion coefficients and interface currents of the two-node NEM need not be chosen as the solution variables to evaluate the residual functions of the NEM_TNCMFD_JFNK method, therefore, the NEM_TNCMFD_JFNK method can greatly reduce the number of solution variables and the computational cost compared with the JFNK based on the conventional NEM. Finally the NEM_TNCMFD_JFNK code is developed and then analyzed by simulating the representative PWR MOX/UO₂ core benchmark, the popular NEACRP 3D core benchmark and the complicated full-core pin-by-pin homogenous core model. Numerical solutions show that the proposed NEM_TNCMFD_JFNK method with the MILU preconditioner has the good numerical accuracy and can obtain higher computational efficiency than the NEM-based two-node CMFD algorithm with the power method in the outer iteration and the Krylov method using the MILU preconditioner in the inner iteration, which indicates the NEM_TNCMFD_JFNK method can serve as a potential and efficient numerical tool for reactor neutron diffusion analysis module in the JFNK-based multiphysics coupling application.