• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.603-616
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• 2007

Roy Huddle, having invented the coated particle in Harwell 1957, stated in the early 1970s that we know now everything about particles and coatings and should be going over to deal with other problems. This was on the occasion of the Dragon fuel performance information meeting London 1973: How wrong a genius be! It took until 1978 that really good particles were made in Germany, then during the Japanese HTTR production in the 1990s and finally the Chinese 2000-2001 campaign for HTR-10. Here, we present a review of history and present status. Today, good fuel is measured by different standards from the seventies: where $9*10^{-4}$ initial free heavy metal fraction was typical for early AVR carbide fuel and $3*10^{-4}$ initial free heavy metal fraction was acceptable for oxide fuel in THTR, we insist on values more than an order of magnitude below this value today. Half a percent of particle failure at the end-of-irradiation, another ancient standard, is not even acceptable today, even for the most severe accidents. While legislation and licensing has not changed, one of the reasons we insist on these improvements is the preference for passive systems rather than active controls of earlier times. After renewed HTGR interest, we are reporting about the start of new or reactivated coated particle work in several parts of the world, considering the aspects of designs/ traditional and new materials, manufacturing technologies/ quality control quality assurance, irradiation and accident performance, modeling and performance predictions, and fuel cycle aspects and spent fuel treatment. In very general terms, the coated particle should be strong, reliable, retentive, and affordable. These properties have to be quantified and will be eventually optimized for a specific application system. Results obtained so far indicate that the same particle can be used for steam cycle applications with $700-750^{\circ}C$ helium coolant gas exit, for gas turbine applications at $850-900^{\circ}C$ and for process heat/hydrogen generation applications with $950^{\circ}C$ outlet temperatures. There is a clear set of standards for modem high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a $500{\mu}m$ diameter $UO_2$ kernel of 10% enrichment is surrounded by a $100{\mu}m$ thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of $35{\mu}m$ thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum $1600^{\circ}C$ afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modem coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond $1600^{\circ}C$ for a short period of time. This work should proceed at both national and international level.

• Journal of the Korean Vacuum Society
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• v.18 no.4
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• pp.318-330
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• 2009

A high heat flux test facility using a graphite heating panel was constructed and is presently in operation at Korea Atomic Energy Research Institute, which is called KoHLT-1. Its major purpose is to carry out a thermal cycle test to verify the integrity of a HIP (hot isostatic pressing) bonded Be mockups which were fabricated for developing HIP joining technology to bond different metals, i.e., Be-to-CuCrZr and CuCrZr-to-SS316L, for the ITER (International Thermonuclear Experimental Reactor) first wall. The KoHLT-1 consists of a graphite heating panel, a box-type test chamber with water-cooling jackets, an electrical DC power supply, a water-cooling system, an evacuation system, an He gas system, and some diagnostics, which are equipped in an authorized laboratory with a special ventilation system for the Be treatment. The graphite heater is placed between two mockups, and the gap distance between the heater and the mockup is adjusted to $2{\sim}3\;mm$. We designed and fabricated several graphite heating panels to have various heating areas depending on the tested mockups, and to have the electrical resistances of $0.2{\sim}0.5$ ohms during high temperature operation. The heater is connected to an electrical DC power supply of 100 V/400 A. The heat flux is easily controlled by the pre-programmed control system which consists of a personal computer and a multi function module. The heat fluxes on the two mockups are deduced from the flow rate and the coolant inlet/out temperatures by a calorimetric method. We have carried out the thermal cycle tests of various Be mockups, and the reliability of the KoHLT-1 for long time operation at a high heat flux was verified, and its broad applicability is promising.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.9-17
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• 2007

In order to compare the correlation of radioactivity ratio between the radwaste streams and the primary coolant of PWR NPPs, A RCS sampling kit was installed to primary coolant system for the collection of the radionuclides during the normal operation of NPPs. RCS samples were collected from PWR type of domestic NPPs through 2004 to 2005, and pretreated with acid microwave digestion or leaching method to assay quantitatively of several interesting radionuclides. The radioactivity ratios of $^{137}Cs\;to\;^{60}Co$ in a filter cartridge and a resin cartridge were 2.3E-2 and 7.3E-1, respectively. At a same period of the reactor operating cycle, the radioactivity ratios of $^{137}Cs\;to\;^{60}Co$ were 6.3E-1 for a evaporator bottom, 6.7E-1 for a spent resin, and 5.6E-2 for a dry active waste, so that these radwaste streams were identified as having similar characteristics with the corresponding RCS samples.

• Journal of Korean Traditional Oncology
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• v.16 no.1
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• pp.41-53
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• 2011

Background and Objectives: Iodine is an essential constituent of the thyroid hormones associated with the growth and development of humans and animals as an inorganic nutrition. This element may be accumulated in human blood, tissues and body through the intake of foodstuffs, a beverage, a nutritional supplement and a medicine, among others. The aim of the research is to find out a better medicinal stuff for the thyroid cancer patient who required a low level of iodine diet. Methods: Neutron activation analysis (NAA) used for the iodine analysis is one of nuclear analytical techniques using radiation and radioisotopes and very useful as sensitive analytical technique for performing both qualitative and quantitative multi-elemental non-destructive analysis of major, minor and trace components in variety of environmental and biological materials. In this study, iodine contents in ten kinds of oriental herb medicinal products, which is frequently used to cancer patients are determined by using instrumental neutron activation analysis (INAA) at the HANARO research reactor. The samples prescribed are manufactured as powdered form for taking medicine easily. The analytical quality control is performed to assure an uncertainty of the measurement and to compensate the measured data using a biological certified reference material, NIST SRM 1572, Citrus Leaves. The measured value is $1.89{\pm}0.35mg/kg$, and the relative error is 2.88%, and relative standard deviation is 19 % due to high counting error by small counts of gamma ray spectrum. The standard deviations for other elements such as Cl, K, Mn and Na were in the range of 2 to 8%. Result: The level of iodine contents of Biki-huan, Chungryong-huan and Chungcho-huan, samples detected is less than 6 mg/kg except Hangam Plus sample (more than 210 mg/kg) and six samples were not detected. Iodine in the samples of Shoxiho-tang, Shopunghualhyl-tang, Shocungryong-tang, Banhasaxim-tang, Insampaedox-san and Myunyuk Plus were not measured, but possible level of content can be estimated from the detection limits. In addition, the concentrations of some major elements like Cl, K, Mn, Na, in the samples were determined with the detection limits. Conclusions: Most of samples showed low iodine contents of less than 6 mg/kg but it turned out that most of testing samples can be used to classify the level of iodine diet samples considering the recommended low level of iodine diet 50 ${\mu}g$/day, and a better medicinal stuff for the thyroid cancer patient can be found.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.105-112
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• 2005

The electrolytic reduction of uranium oxide in a LiCl-Li$_{2}$O molten salt system has been studied in a 10 g U$_{3}$O$_{8}$ /batch-scale experimental apparatus with an integrated cathode assembly at 650$^{\circ}C$. The integrated cathode assembly consists of an electric conductor, the uranium oxide to be reduced and the membrane for loading the uranium oxide. From the cyclic voltammograms for the LiCl-3 wt$\%$ Li$_{2}$O system and the U$_{3}$O$_{8}$-LiCl-3 wt$\%$ Li$_{2}$O system according to the materials of the membrane in the cathode assembly, the mechanisms of the predominant reduction reactions in the electrolytic reactor cell were to be understood; direct and indirect electrolytic reduction of uranium oxide. Direct and indirect electrolytic reductions have been performed with the integrated cathode assembly. Using the 325-mesh stainless steel screen the uranium oxide failed to be reduced to uranium metal by a direct and indirect electrolytic reduction because of a low current efficiency and with the porous magnesia membrane the uranium oxide was reduced successfully to uranium metal by an indirect electrolytic reduction because of a high current efficiency.

• Journal of Hydrogen and New Energy
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• v.26 no.2
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• pp.120-126
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• 2015

Recently, the ever-increasing use of fossil fuels for rapid industrial development and population significantly caused an environment pollution and global warming such as climate change. So research and development of sustainable and eco-friendly energy have been performed. Especially the interest in nuclear fusion fuel was significantly increased from the developed countries. The system of fusion energy production was tritium separation, storage and delivery, and purification. Republic of Korea is in charge of Storage and Delivery System (SDS) in the International Thermonuclear Experimental Reactor (ITER). Welding part of the SDS bottles for storing the tritium is known to be susceptible to hydrogen embrittlement. In this study, conducted a study for the relaxation of the stability and hydrogen embrittlement of the weld area. The hydrogen heat treatment was processed through the Pressure-Composition-Temperature (PCT) device according to the time variation. Also mechanical properties such as impact test and hardness test according to using the alkaline cleaning liquid for hydrogen embrittlement relief and the fracture was observed by scanning electron microscopy (SEM) after the mechanical properties evaluation.

• The Korean Journal of Nuclear Medicine
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• v.34 no.1
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• pp.62-73
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• 2000

Purpose: Esophageal cancer patients have a difficulty in the intake of meals through the blocked esophageal lumen, which is caused by an ingrowth of cancer cells and largely influences on the prognosis. It is reported that esophageal cancer has a very low survival rate due to the lack of nourishment and immunity as the result of this. In this study a new radioactive stent, which prevents tumor ingrowth and restenosis by additional radiation treatment, has been developed. Materials and Methods: Using ${\ulcorner}HANARO{\lrcorner}$ research reactor, the radioactive stent assembly ($^{166}Ho$-SA) was prepared by covering the metallic stent with a radioactive sleeve by means of a post-irradiation and pre-irradiation methods. Results: Scanning electron microscopy and autoradiography exhibited that the distribution of $^{165/166}Ho\;(NO_3)$ compounds in polyurethane matrix was homogeneous. A geometrical model of the esophagus considering its structural properties, was developed for the computer simulation of energy deposition to the esophageal wall. The dose distributions of $^{166}Ho$-stent were calculated by means of the EGS4 code system. The sources are considered to be distributed uniformly on the surface in the form of a cylinder with a diameter of 20 mm and length of 40 mm. As an animal experiment, when radioactive stent developed in this study was inserted into the esophagus of a Mongrel dog, tissue destruction and widening of the esophageal lumen were observed. Conclusion: We have developed a new radioactive stent comprising of a radioactive tubular sleeve covering the metallic stent, which emits homogeneous radiation. If it is inserted into the blocked or narrowed lumen, it can lead to local destruction of the tumor due to irradiation effect with dilatation resulting from self-expansion of the metallic property. Accordingly, it is expected that restenosis esophageal lumen by the continuous ingrowth and infiltration of cancer after insertion of our radioactive stent will be decreased remarkably.

• Nuclear Engineering and Technology
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• v.1 no.2
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• pp.91-101
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• 1969

Resistivity changes of n-type float-zone silicon crystals with 6.4$\times$10$^{14}$ to 1.25$\times$10$^{17}$ phosphorus atoms/㎤ due to irradiation by (1) 1 MeV electrons, (2) two types of research reactors, and (3) $Co^{60}$ ${\gamma}$-ray sources were investigated. The results were analyzed on the basis of a simple exponential formula derived by Buehler. While the formula gave a fair fit in the low fluence range in most cases, the deviation was quite appreciable in the case of 1 MeV electron irradiation, and a linear change gave better fit in some cases. The large change in the carrier removal rate in electron-irradiated samples in the high fluence range was analyzed in detail in terms of the Fermi level cross-over of the defect levels. Based on the damage constants evaluated from the initial portion of data where the formula was applicable, the relative effectiveness of various radiation sources in causing the resistivity change in n-type silicon was compared. The TRIGA Mark II reactor neutrons, for example, were found to be about 40 times more effective than 1 MeV electrons. The dependence of the damage constant on the initial carrier concentration was also examined. The physical basis of the exponential law and the effect of the Fermi level cross-over of the defect levels on the resistivity change in the high fluence ranges are discussed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.421-429
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• 2018

Hydrazine based reductive dissolution applied on magnetite oxide was investigated. Dissolution of Fe(II) and Fe(III) from magnetite takes place either by protonation, surface complexation, or reduction. Solution containing hydrazine and sulfuric acid provides hydrogen to break bonds between Fe and oxygen by protonation and electrons for the reduction of insoluble Fe(III) to soluble Fe(II) in acidic solution of pH 3. In terms of dissolution rate, numerous transition metal ions were examined and Cu(II) ion was found to be the most effective to speed up the dissolution. During the cycle of Cu(I) ions to Cu(II) ions, the released electron promoted the reduction of Fe(III) and Cu(II) ions returned to Cu(I) ion due to the oxidation of hydrazine. In the experimental results, the addition of a very low amount of cupric ion (about 0.5 mM) to the solution increased the dissolution rate about 40% on average and up to 70% for certain specific conditions. It is confirmed that even though the coordination structure of copper ions with hydrazine is not clear, the $Cu(II)/H^+/N_2H_4$ system is acceptable regarding the dissolution performance as a decontamination reagent.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.95-104
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• 2005

Effects of the aluminum melting temperature, melting time and a kind of flux agents on the distribution of surrogate nuclide were investigated in the electric furnace at the aluminum melting including surrogate radionuclides(Co, Cs, Sr) in order to establish the fundamental research of the melting technology for the metallic wastes from the decommissioning of the TRIGA research reactor. It was verified that the fluidity of aluminum melt was increased by adding flux agent but it was slightly varied according to the sort of flux agents. The results of the XRD analysis showed that the surrogate nuclides move into the slag phase and then they were combined with aluminum oxide to form more stable compound. The weight of the slag generated from aluminum melting test increased with increasing melting temperature and melting time and the increase rate of the slag depended on the kind of flux agents added in the aluminum waste. The concentration of the cobalt in the ingot phase decreased with increasing reaction temperature but it increased in the slag phase up to 90$\%$according to the experimental conditions. The volatile nuclides such as Cs and Sr considerably transferred from the ingot phase to the slag and dust phase.