• Nuclear Engineering and Technology
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• v.54 no.8
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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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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• v.53 no.10
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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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• v.54 no.3
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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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• v.49 no.7
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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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• v.32 no.4
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• pp.303-310
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• 2002

Spermatogenesis and sperm ultrastructure are investigated by means of light and transmission electron microscopy in the equilateral venus, Gomphina veneriformis which is dominant bivalve in the east coast of Korea. In the active spermatogenic season, testis consists of numerous spermatogenic follicles which is contains germ cells in the different developmental stage. The spermatogonia attached to spermatogenic follicle wall and has a large nucleus with electron-dense nucleolus. The spermatocytes are characterized by appearance of synaptonemal complex and well-developed Golgi complex. Nucleus of spermatid consists of numerous heterogeneous granules with high electron density. Karyoplasmic condensation, acrosome and flagellum formations are observed during spermiogenesis. Testicular matured sperms of sperm bundle consists of head, midpiece and tail. The head is about $8.5{\mu}m$ long and comprises a long nucleus and a bullet-like acrosome ($8.5{\mu}m$ in length). Acrosomal rod of microfilaments is observed in the lumen between nucleus and acrosome. The midpiece has four mitochondria. And tail has the typical '9+2' microtubule system.

• Journal of fish pathology
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• v.7 no.1
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• pp.1-5
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• 1994

Nuclear polyhedrosis virus was successfully infected the continuous Sf cell line. At 12hrs post-infectio(P.I), the cell lost the motility and the nuclei of the cells were hypertrophied. At 24hrs P.I, the cells were somewhat abnormal form and PIB formation was observed. At 48hrs, the PIBs formed in all cells. PIBs in the nuclei were released in the culture media at 72hrs P.I. By the observation of NPV morphogenesis by electron microscopy at 13hrs P. I, the virogenic stroma formed in the nucleus, and nucleocapsids formed. At 48hrs P.I, many nucleocapsids were bundled and then occluded in PIB, and PIBs were matured. PIB shapes were mostly tetragonal and a polyhedron was about $3{\sim}10{\mu}m$ in size. Virions were rod shape. nucleocapsids ranging in size $30{\sim}40{\times}300{\sim}400nm$.

• Journal of fish pathology
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• v.21 no.1
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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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• v.15 no.1
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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.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.166-177
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• 1993

A simplified thermal analysis method to evaluate the maximum temperature of the CANDU 37-element fuel bundle within a fuel basket in a given spent fuel dry storage canister has been presented along with the results of sample analyses performed to examine the parametric effects of the ambient conditions on the maximum fuel temperature within a canister. To solve the multi-dimensional heat transfer problem of the complex geometry of rod bundles within a canister where three modes of heat transfer are superimposed, the CANDU spent fuel bundles stored in the dry storage canister are first replaced by equivalent concentric fuel cylinders. The simplified axi-symmetric two-dimensional multi-mode heat transfer problem of the equivalent fuel cylinders is then analyzed with an existing computer code, HEATING5, using additional input data and heat transfer correlations. A comparison between the predicted temperature profile and the mock-up test results shows that the agreement is quite satisfactory.