• 한국세라믹학회지
• /
• 제44권10호
• /
• pp.580-584
• /
• 2007

The ZrC layer instead of SiC layer is a critical and essential layer in TRISO coated fuel particles since it is a protective layer against diffusion of fission products and provides mechanical strength for the fuel particle. In this study, we carried out computational simulation before actual experiment. With these simulation results, Zirconium carbide (ZrC) films were chemically vapor deposited on $ZrO_2$ substrate using zirconium tetrachloride $(ZrCl_4),\;CH_4$ as a source and $H_2$ dilution gas, respectively. The change of input gas ratio was correlated with growth rate and morphology of deposited ZrC films. The growth rate of ZrC films increased as the input gas ratio decreased. The microstructure of ZrC films was changed with input gas ratio; small granular type grain structure was exhibited at the low input gas ratio. Angular type structure of increased grain size was observed at the high input gas ratio.

• Nuclear Engineering and Technology
• /
• 제38권6호
• /
• pp.459-482
• /
• 2006

A geochemical approach to the long-term safety of waste disposal is discussed in connection with the significance of actinides, which shall deliver the major radioactivity inventory subsequent to the relatively short-term decay of fission products. Every power reactor generates transuranic (TRU) elements: plutonium and minor actinides (Np, Am, Cm), which consist chiefly of long-lived nuclides emitting alpha radiation. The amount of TRU actinides generated in a fuel life period is found to be relatively small (about 1 wt% or less in spent fuel) but their radioactivity persists many hundred thousands years. Geological confinement of waste containing TRU actinides demands, as a result, fundamental knowledge on the geochemical behavior of actinides in the repository environment for a long period of time. Appraisal of the scientific progress in this subject area is the main objective of the present paper. Following the introductory discussion on natural radioactivities, the nuclear fuel cycle is briefly brought up with reference to actinide generation and waste disposal. As the long-term disposal safety concerns inevitably with actinides, the significance of the aquatic actinide chemistry is summarized in two parts: the fundamental properties relevant to their aquatic behavior and the geochemical reactions in nanoscopic scale. The constrained space of writing allows discussion on some examples only, for which topics of the primary concern are selected, e.g. apparent solubility and colloid generation, colloid-facilitated migration, notable speciation of such processes, etc. Discussion is summed up to end with how to make a geochemical approach available for the long-term disposal safety of nuclear waste or for the performance assessment (PA) as known generally.

• 한국미생물·생명공학회지
• /
• 제23권3호
• /
• pp.255-262
• /
• 1995

Intact cells of Acinetobacter sp. T5-7 completely degraded trichloroethylene (TCE) following growth with phenol. This strain could grow on at least eleven aromatic compounds, e.g., benzaldehyde, benzene, benzoate, benzylalochol, catechol, caffeic acid, 2.4-D, p-hydroxybenzoate, phenol, protocatechuate and salicylate, and did grow on alkane, such as octane. But except phenol, other aromatic compounds did not induced TCE degradation. Phenol biotransformation products, catechol was identified in the culture media. However, catechol-induced cells did not degrade TCE. So we assumed that phenol hydroxylase was responsible for the degradation of TCE. The isolate T5-7 showed growth in MM2 medium containing sodium lactate and catechol rather than phenol, but did not display phenol hydroxyalse activity, suggesting induction of enzyme synthesis by phenol. Phenol hydroxylase activity was independent of added NADH and flavin adenine dinucleotide but was dependent on NADPH addition. Degradation of phenol produced catechols which are then cleaved by meta-fission. We identified catechol-2.3-dioxygenase by active staining of polyacrylamide gel.

• 패션비즈니스
• /
• 제12권6호
• /
• pp.23-33
• /
• 2008

The physico-chemical characteristics by bleaching treatments were assessed by several instrumental analyses such as surface morphology, chemical structural change, color change as well as tensile strength. The change of morphological characteristic was observed through scanning electron microscope(SEM). The observation of the fine structure on hair surface by SEM showed the bleached hair had much damaged to hair cuticle, and some of cuticle surface were worn away. To investigate the chemical structural changes in hair keratin, the cross-sections of hair samples were directly analysed using Fourier transform infrared microspectroscopy(FT-IRM). The results showed the cysteic acid S=O band intensity was distinctively increased by performing the bleaching treatment. The cleavage of cystine was appeared to proceed primarily through the sulfur-sulfur (-S-S-) fission whereby cysteic acid was formed as a principal oxidation products. The distribution of amide I band in hair keratin was determined by attenuated total reflectance(ATR) FT-IR mapping image. The results showed that the outer side of hair cortex was more damaged than the inner side of the hair cortex. Also, during chemical bleaching of the hair with alkaline peroxide, the hair was turned to reddish yellow due to the oxidative degradation of eumelanin. This means the eumelanin is more unstable than pheomelanin in chemical oxidation. With bleaching, the tensile strength was also reduced as a results of the chemical oxidation.

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
• /
• pp.176-181
• /
• 2005

The electrolytic reduction process and the electrorefining process, which are being developed at the Korea Atomic Energy Research Institute (KAERI), are to generate molten waste salts such as LiCI salt and LiCI-KCI eutectic salt, respectively. Our goal in waste salt management is to minimize a total waste generation and fabricate a very low­leaching waste form such as a ceramic waste form. Zeolite has been known to one of the most desirable media to immobilize waste salt, which is water soluble and easily radiolyzed. Zeolite can be also used to the removal of fission products from the spent waste salt. Molten LiCI salt is mixed with zeolite A at $650^{\circ}C$ to form a salt-loaded zeolite, and then thermally treated in above $900^{\circ}C$ to become an immobilized product with crystal phase of $Li_{8}Cl_{2}$-Sodalite. In this work, a crystal phase changes of immobilization medium, zeolite, during immobilization of molten LiCI salt using zeolite A is introduced.

• 한국압력기기공학회 논문집
• /
• 제15권2호
• /
• pp.64-70
• /
• 2019

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

• 한국방사성폐기물학회:학술대회논문집
• /
• 한국방사성폐기물학회 2003년도 가을 학술논문집
• /
• pp.406-410
• /
• 2003

DUPIC 핵연료 제조시설은 PWR 사용후핵연료를 건식 재가공하여 CANDU 형 핵연료를 제조하는 시설이다. DUPIC 시설과 같이 사용후핵연료를 취급하는 시설에서 핵물질 안전조치를 위해 적용되는 연속 무인 감시시스템은 많은 양의 영상 및 방사선 감시 데이터를 생산하게 되며, 이러한 자료로부터 핵물질의 전용 여부를 분석하기 위해서는 상당한 시간과 인력이 소요된다. 따라서 핵물질 취급시설에서의 감시시스템은 시설로부터 취득한 감시 데이터를 자동적으로 검토ㆍ분석하여 비정상적인 상황을 추출해 낼 수 있는 기능이 요구된다. 이 연구에서는 이러한 관점에서 영상 및 방사선 데이터를 자동 분석할 수 있는 신경망을 이용한 지능형 핵물질 감시시스템을 개발하였다. DUPIC 시설의 안전조치를 위해 개발한 동 핵물질 감시시스템은 수차례의 성능 시험을 거쳐, 현재 시설에 설치되어 정상적으로 운영 중에 있다.

• Nuclear Engineering and Technology
• /
• 제31권1호
• /
• pp.68-79
• /
• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• Nuclear Engineering and Technology
• /
• 제40권7호
• /
• pp.581-592
• /
• 2008

In this study, a conceptual pyroprocess flowsheet has been devised by combining several dry-type unit processes; its applicability as an alternative fuel cycle technology was analyzed. A key point in the evaluation of its applicability to the fuel cycle was the recovery yield of fissile materials from spent fuels as well as the proliferation resistance of the process. The recovery yields of uranium and transuranic elements (TRU) were obtained from a material balance for every unit process composing the whole pyroprocess. The material balances for several elemental groups of interest such as uranium, TRU, rare earth, gaseous fission products, and heat generating elements were calculated on the basis of the knowledge base that is available from domestic and foreign experimental results or technical information presented in open literature. The calculated result of the material balance revealed that uranium and TRU could be recovered at 98.0% and 97.0%, respectively, from a typical PWR spent fuel. Furthermore, the anticipated TRU product was found to emit a non-negligible level of $\gamma$-ray and a significantly higher level of neutrons compared to that of a typical plutonium product obtained from the PUREX process. The results indicate that the product from this conceptual pyroprocessing should be handled in a shielded cell and that this will contribute favorably to retaining proliferation resistance.

• Nuclear Engineering and Technology
• /
• 제48권1호
• /
• pp.1-15
• /
• 2016

For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell $UO_2$ and high-density composite pellet concepts are being developed as ATF pellets. A microcell $UO_2$ pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts-surface-modified Zr-based alloy and SiC composite material-are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.