• Nuclear Engineering and Technology
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• 제37권1호
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• pp.79-90
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• 2005

In this paper are reviewed the current status of nuclear fuel management and reactor operational aid tools. In addition, we indicate deficiencies in current capabilities and what future research is judged warranted. For the nuclear fuel management review the focus is on light water reactors and the utilization of stochastic optimization methods applied to the lattice, fuel bundle, core loading pattern, and for BWRs the control rod pattern/core flow design decision making problems. Significant progress in addressing separately each of these design problems on a single cycle basis is noted; however, the outstanding challenge of addressing the integrated design problem over multiple cycles under conditions of uncertainty remains to be addressed. For the reactor operational aid tools review the focus is on core simulators, used to both process core instrumentation signals and as an operator aid to predict future core behaviors under various operational strategies. After briefly reviewing the current status of capabilities, a more in depth review of adaptive core simulation capabilities, where core simulator input data are adjusted within their known uncertainties to improved agreement between prediction and measurement, is presented. This is done in support of the belief that further development of adaptive core simulation capabilities is required to further significantly advance the utility of core simulators in support of reactor operational aid tools.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• 한국압력기기공학회 논문집
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• 제20권1호
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• pp.32-38
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• 2024

The best estimate plus uncertainty methodologies for loss-of-coolant accident analyses make use of the best-estimate codes and relevant experimental databases. Inherently, best-estimate codes have various uncertainties in the model parameters, which can be quantified by the dedicated experimental database. Therefore, this study was devoted to establishing procedures for identifying the input parameters of predictive models and quantifying their uncertainty ranges. The rod bundle heat transfer experiments were employed as a representative reflood separate effect test, and the TRACE code was utilized as a best-estimate code. In accordance with the present procedure for uncertainty quantification, the integrated list of the influential input parameters and their uncertainty ranges was obtained through local sensitivity calculations and screening criteria. The validity of the procedure was confirmed by applying it to uncertainty analyses, which checks whether the measured data are within computed ranges of the variables of interest. The uncertainty quantification procedure proposed in this study is anticipated to provide comprehensive guidance for the conduct of uncertainty analyses.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.842-848
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• 2022

Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 ℃. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner subchannel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Applied Microscopy
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• 제32권4호
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• pp.303-310
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• 2002

우리나라 동해안에 가장 많이 서식하는 조개류인 대복의 정소구조와 정자형성과정을 광학현미경과 투과전자현미경으로 조사한 결과는 다음과 같다. 대복의 정소는 소성결합조직으로 구성된 다수의 정자형성 소낭을 가진다. 동일한 정자형성 소낭 내에서는 여러 단계의 생식세포들이 관찰되었다. 정원세포들은 정자형성 소낭벽에 부착되어 있으며, 커다란 핵과 뚜렷한 인을 가진다. 정모세포에서는 연접사복합체와 골기체의 발달을 확인할 수 있었다. 정세포의 핵은 전자밀도가 높은 과립상의 염색질로 구성되며, 정자변태과정 동안에 핵의 응축 및 첨체와 편모의 형성을 관찰할 수 있었다. 정소 내에서 완숙 정자들은 다발을 형성하고 있으며, 두부, 중편, 미부로 구성되어 있었다. 두부의 길이는 약 $8.5{\mu}m$로, 첨체부와 핵 부위로 구분된다. 첨체는 길이 약 $1.1{\mu}m$의 총알형태였다. 두부와 첨체 사이에서는 미세섬유로 구성된 첨체기둥이 확인되었다. 중편에는 4개의 미토콘드리아를 가지며, 꼬리의 횡단면은 "9+2"의 구조를 나타냈다.

• 한국어병학회지
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• 제7권1호
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• pp.1-5
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• 1994

핵다면체 바이러스는 S. frugiferda 세포주에 감염되었다. 감염 12시간 후에 세포는 운동성을 잃고 세포의 핵은 팽창되었다. 감염 24시간 후에 세포는 비정상적인 형태로 되었으며, 작은 PIB가 나타나기 시작하였다. 감염 48시간 후에는 전체 세포에 PIB가 형성되었고, 감염 76시간 후에는 핵속에 있는 PIB는 일부가 세포밖으로 방출되었다. 전자현미경을 통해 감염후 13시간이 지난후 NPV를 관찰한 결과 세포의 핵속에서는 virogenic stroma가 형성되었으며, 그 부분에서 nucleocapsids가 형성되었다. 감염 48시간 후에는 많은 nucleocapsid들은 bundle을 형성하고, PIB에 봉입되었다. PIB는 대부분 4면체이며, 크기는 $3{\sim}10{\mu}m$ 정도이었다. Virion은 막대형으로 nucleocapsid는 $30{\sim}40{\times}300{\sim}400nm$이었다.

• 한국어병학회지
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• 제21권1호
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• pp.67-79
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• 2008

This study was conducted to find out biological response of bivalves exposed to tributyltin chloride(TBTCl). The results of the study confirmed that TBTCl induce the reduction of oxygen consumption rateand histopathological feature in the gill structure of equilateral venus, Gomphina veneriformis. The experi-mental groups consisted of a control and 3 TBTCl exposure groups (0.4, 0.6 and 0.8 yg TBTCl L') and theexperimental period was 36 weeks. For histological analysis, gill tissues were fixed in Bouin's fluid andthen stained H-E stain, AB-PAS (PH 2.5) reaction and Masson's trichrome stain after having serially sec-tioned the tissue by paraffin method at thickness of 4-6 (an. The oxygen consumption rate was not signifi-cantly different between the control and exposure groups at 4 weeks, but in all exposure groups at 28 weeks,it was significantly different to the control. Gill of G. veneriformis had demibranch that attached two sheetsof lamellae and a lamella was composed of numerous filaments, numbering 25 on average. The frontal fila-ment zone had three types of cilia; frontal, latero-frontal and lateral depending on locations while the lateralcilia were the longest and largest in number. The mucous cells observed in filaments were more abundant in(542c) in AB-PAS (PH 2.5) reaction. Gill exposed to TBTCl was extended hemolymph sinus and increased hemocytes at 4 weeks, and then it showed increases of mucous cells and partially disappearance of frontalcilia. In the group of 0.8 yg TBTCl L' at 12 weeks, hypertrophy of frontal and latero-frontal epithelia wasobserved. Also it observed m decrease of mucous cell containing weekly acid mucosubstance and appearedpartially destruction muscle fiber bundle, In the groups of 0.4 and 0.6 ug TBTCl L' at 36 weeks, it appearedpartially modification of epithelia and in 0.8 us TBTCl L' group, observed filaments that come out chiti-nous rod from disappearance of frontal and latero-frontal epithelia.

• Applied Microscopy
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• 제15권1호
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• pp.71-85
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• 1985

The ultrastructure on the dorsal vessel of 5-day-old cabbage butterfly, Pieris rapae L., was carried out using the transmission and scanning electron microscope. The results are as follows. 1) The aorta. The aorta is simple tubular type and consists of the inner and outer membrane of the myocardium and thick myocardium is located between them. However the inner membrane with $0.26{\mu}m$ thickness and outer membrane with $0.08{\mu}m$ are composed of fibrous materials, the former is composed of low and high densed fibrous materials and the latter appears homogeneous layer. The myocardium consists of typical striated muscles. The sarcomere with $1.6{\mu}m$ length and in cross section, each thick filaments are surrounded by $7{\sim}8$ thin filaments. The intercalated disc is joining the end of the two muscle cells, desmosomes and septate junctions are appeared between the neighboring muscle cells. 2) The heart. The heart composing of myocardium enclosed by its inner and outer membrane as the aorta has a series of well formed segmental chamber. The arrangement of myofilaments, cell adhensions and membrane elements are observed as same as at the aorta. The inner membrane of the heart is deeply invaginated into the myocardium than the outer membrane and a lot of well developed mitochondria with rod shape are aggregated in the folds. The longitudinally and transversely oriented tubule system formed by invagnation of the sarcolemma into the muscle bundle is built up dyad with the sarcoplasmic reticulum as the aorta. The slit is formed by deeply invagination of the inner membrane of myocadium toward the muscle layer and then the inner and outer membrane of myocardium are fused. Therefore, the ostium is formed between the myocardium and situated at the lateral side of the myocardium.