• Nuclear Engineering and Technology
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• 제30권6호
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• pp.541-554
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• 1998

A computer code COSMOS has been developed based on the CARO-D5 for the thermal analysis of LWR UO$_2$ and MOX fuel rod under steady-state and transient operating conditions. The main purpose of the COSMOS, which considers high turnup characteristics such as thermal conductivity degradation with turnup and rim formation at the outer part of fuel pellet, is to calculate temperature profile across fuel pellet and fission gas release up to high burnup. A new mechanistic fission gas release model developed based on physical processes has been incorporated into the code. In addition, the features of MOX fuel such as change in themo-mechanical properties and the effect of microscopic heterogeneity on fission gas release have been also taken into account so that it can be applied to MOX fuel. Another important feature of the COSMOS is that it can analyze fuel segment refabricated from base irradiated fuel rods in commercial reactors. This feature makes it possible to analyze database obtained from international projects such as the MALDEN and RISO, many of which were collected from refabricated fuel segments. The capacity of the COSMOS has been tested with some number of experimental results obtained from the HALDEN, RISO and FIGARO programs. Comparison with the measured data indicates that, although the COSMOS gives reasonable agreement, the current models need to be improved. This work is being performed using database available from the OECD/NEA.

• Corrosion Science and Technology
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• 제20권4호
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2359-2366
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• 2022

Selecting graphite grades with superior irradiation characteristics is important task for designers of graphite moderation reactors. To provide reference information and data for graphite selection, the effects of irradiation on three fine-grained, iso-molded nuclear grade graphites, ETU-10, IG-110, and NBG-25, were compared based on irradiation-induced changes in volume, thermal conductivity, dynamic Young's modulus, and coefficient of thermal expansion. Data employed in this study were obtained from reported irradiation test results in the high flux isotope reactor (HFIR)(ORNL) (ETU-10, IG-110) and high flux reactor (HFR)(NRL) (IG-110, NBG-25). Comparisons were made based on the irradiation dose and irradiation temperature. Overall, the three grades showed similar irradiation-induced property change behaviors, which followed the historic data. More or less grade-sensitive behaviors were observed for the changes in volume and thermal conductivity, and, in contrast, grade-insensitive behaviors were observed for dynamic Young's modulus and coefficient of thermal expansion changes. The ETU-10 of the smallest grain size appeared to show a relatively smaller VC to IG-110 and NBG-25. Drastic decrease in the difference in thermal conductivity was observed for ETU-10 and IG-110 after irradiation. The similar irradiation-induced properties changing behaviors observed in this study especially in the DYM and CTE may be attributed to the assumed similar microstructures that evolved from the similar size coke particles and the same forming method.

• 방사성폐기물학회지
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• 제21권4호
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2008년도 춘계학술대회 논문집
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• pp.210-213
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• 2008

The purpose of this study is to investigate the performances of organic removal and methane recovery in the full scale two-phase anaerobic system. The full scale two-phase anaerobic system was consists of an acidogenic ABR (Anaerobic Baffled Reactor) and a methanognic UASB (Upflow Anaerobic Sludge Blanket) reactor. The volume of acidogenic and methanogenic reactors is designed to 28.3 $m^3$ and 75.3 $m^3$. The two-phase anaerobic system represented 60-82% of COD removal efficiency when the influent COD concentration was in the range of 7,150 to 16,270 mg/L after screening (average concentration is 10,280 mg/L). After steady-state, the effluent COD concentration in the methanogenic reactor showed 2,740 $\pm$ 330 mg/L by representing average COD removal efficiency was 71.4 $\pm$ 8.1% when the operating temperature was in the range of 19-32$^{\circ}C$. The effluent SCOD concentration was in the range of 2,000-3,000 mg/L at the steady state while the volatile fatty concentration was not detected in the effluent. Meanwhile, the COD removal efficiency in the acidogenic reactor showed less than 5%. The acidogenic reactor played key roles to reduce a shock-loading when periodic shock loading was applied and to acidify influent organics. Due to the high concentration of alkalinity and high pH in the effluent of the methanogenic reactor, over 80% of methane in the biogas was produced consistently. More than 70 % of methane was recovered from theoretical methane production of TCOD removed in this research. The produced gas can be directly used as a heat source to increase the reactor temperature.

• 유기물자원화
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• 제23권3호
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• pp.51-59
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• 2015

가축분뇨 슬러지는 유기물의 농도가 높고 일부 중금속이 높은 농도로 혼합되어 있어 해양에 배출될 경우 환경에 부정적인 영향을 끼칠 수 있다. 본 연구에서는 가축분뇨 슬러지 처리에 열가수분해 기술을 적용하여 연료화 가능성을 판단하고 최적 운전조건을 도출하고자 하였다. 밀폐형 고압반응기를 사용하여 가축분뇨 슬러지를 $170{\sim}210^{\circ}C$까지 온도변화를 주면서 열가수분해 하였고, 반응 후 생성된 액상생성물과 탈수케이크의 분석을 실시하였다. 반응온도 $190^{\circ}C$로 운전하는 것이 가장 효과적인 것으로 나타났으며, 반응온도 $190^{\circ}C$일 때 고체생성물의 고위발열량은 5,050 kcal/kg, 저위발열량은 4,740 kcal/kg으로 연료로서 충분히 가치가 있는 것으로 판단되었다.

• 수산해양기술연구
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• 제28권1호
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• pp.71-78
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• 1992

Zr-4 used for a cladding and an end plug of reactor component has creep deformation under operation at high temperature. Creep is regarded as the time dependent deformation of a material under constant applied stress. Although the major source of the deformation of zirconium component in water-cooled reactors is irradiation creep, the thermal creep may give a rise to significant deformation in reactor component especially at relatively high temperatures and at various constant stresses, and therefore it must be predicted accurately. Stress relaxation is the time dependent change of stress at constant strain and it is a process related intimately to creep. In this paper, the creep behavior and stress relaxation of Zr-4 is examined at the temperature of 50$0^{\circ}C$ that is 40% of the absolute melting temperature of Zr-4 under the stress below yield stress and under the various constant strains. The results obtained are summarized as follows: 1) With an increase of stress, the steady state creep rate increases and the creep rupture time decreases. 2) The steady state creep rate $\varepsilon$(%/s) for the stress $\sigma$sub(c) (kgf/mm super(2)) of Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 are in accord with Norton's model equation($\varepsilon$=K$\sigma$ sub(c) super (n)). The constants of materials computed are as follows: K=3.9881$\times$10 super(-5), n=1.9608 3) The rupture time T sub(r) (hr) decreases linearly with the increase of stress on the log-log scaled graph. The empirical equations computed for Zr-4 are in accord with Bailey's model equation (T sub(r)=K sub(1)$\sigma$sub(c) super(m)). The constants of materials computed are as follows: K sub(1)=1.2875$\times$10 super(16), m=-3.467 4) It seems clear that the strain could be quantitatively dependent on the high temperature creep properties such as creep stress, rupture time, steady state creep rate and total creep rate. It is found that these relationships are linear on the log-log graph. 5) In stress relaxation test, as the critical constant strain that can be allowed to the specimen is larger, stress relaxation becomes more rapid, and as the constant strain is smaller, the stress relaxation becomes slower.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• 한국환경농학회지
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• 제17권3호
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• pp.260-267
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• 1998

밀폐형 bench-scale reactor(242 l)에 thermocouples, oxygen sensor 및 datalogger 등을 연결하여 우분과 볏짚 혼합물의 퇴비화 촉진 및 양질의 퇴비생산을 위한 기초자료를 얻고자 공기주입량(0.09, 0.18, 0.90, 1.79 l $min^{-1}kg$ dry $solids^{-1}$)에 따른 생물학적 활성 변수들의 일시적이고 공간적인 변화를 모니터링한 결과는 다음과 같았다. 공기주입량이 높을수록 퇴비화 초기 및 뒤집기 이후 단계 모두 부숙온도의 증가 및 감속속도가 빨랐고 퇴적물과 배출공기의 온도차이는 적었으며, 모든 처리에서 퇴적물의 경우 $50{\sim}53^{\circ}C$에서 약 5시간 동안, 배출공기는 $45^{\circ}C$에서 약 $5{\sim}15$시간 동안 온도정지기를 보인 후 다시 증가하는 경향을 보였다. 최대온도는 퇴비화 초기단계에서는 공기주입량이 많을수록 감소하였지만 뒤집기 이후 단계에서는 1.79 l $min^{-1}kg^{-1}$처리를 제외하고는 그 반대 경향이었으며, 퇴적물의 최고온도 도달시간은 초기 및 뒤집기 이후 단계 모두 공기주입량이 적을수록 늦었다. $45^{\circ}C$이상의 고온 유지시간은 공기주입량이 증가할수록 급격히 감소하였고, 퇴비화 초기단계에 있어서 0.09 및 0.18 l $min^{-1}kg^{-1}$ 처리는 $65^{\circ}C$ 이상, 0.90 l $kg^{-1}$는 $55{\sim}64.9^{\circ}C$, 1.79 l $min^{-1}kg^{-1}$는 $45{\sim}54.9^{\circ}C$의 유지시간이 가장 길었다. 배출공기의 최저 산소농도 및 최대 산소비율은 공기주입량이 많을수록 높아지는 경향이었으나 그 수준에 도달하는 시간은 일정한 경향이 없었다.

• 한국수소및신에너지학회논문집
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• 제25권3호
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• pp.247-254
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• 2014

The synthesis of Fischer-Tropsch oil is the catalytic hydrogenation of CO to give a range of products, which can be used for the production of high-quality diesel fuel, gasoline and linear chemicals. Our cobalt based catalyst was prepared Co/alumina, silica and titania by the incipient wet impregnation of the nitrates of cobalt and promoter with supports. Cobalt catalysts was calcined at $350^{\circ}C$ before being loaded into the FT reactors. After the reduction of catalyst has been carried out under $450^{\circ}C$ for 24hrs, FT reaction of the catalyst has been carried out at GHSV of 4,000/hr under $200^{\circ}C$ and 20atm. From these test results, we have obtained the results as following ; in case of 12wt% Co-supported $Al_2O_3$, $SiO_2$ and $TiO_2$ catalysts, maximum activities of the catalysts were appeared at the promoters of Mn, Mo and Ce respectively. The activity of 12wt% $Co/Al_2O_3$ added a Mn promoter was about 3 times as high as that of 12wt% $Co/Al_2O_3$ catalyst without promoters. When it has been the experiment at the range of reaction temperature of $200{\sim}220^{\circ}C$ and GHSV of 1,546~5,000/hr, the results have shown generally increasing the activities with the increase of reaction temperature and GHSV.