• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2047-2052
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• 2023

Risk-informed approach has been widely applied in the safety design, regulation, and operation of nuclear reactors. It has been commonly accepted that risk-informed design optimization should be used in the innovative reactor designs to make nuclear system highly safe and reliable. In spite of the risk-informed approach has been used in some advanced nuclear reactors designs, such as Westinghouse IRIS, Gen-IV sodium fast reactors and lead-based fast reactors, the process of risk-informed design of nuclear reactors is hardly to carry out when passive system reliability should be integrated in the framework. A practical method for new passive safety reactors based on probabilistic safety assessment (PSA) and passive system reliability analyze linking is proposed in this paper. New three-dimension frequency-consequence curve based on risk concept with three variables is used in this method. The proposed method has been applied to the determination optimization of design options selection in a 10 MWth lead-based research reactor(LR) to obtain one optimized system design in conceptual design stage, using the integrated reliability and probabilistic safety assessment program RiskA, and the computation resources and time consumption in this process was demonstrated reasonable and acceptable.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3571-3580
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• 2023

The application of the induction bending process to pipe systems in various industrial fields is increasing. Recently, efforts have also been made to apply this bending process to nuclear power plants because it can innovatively reduce welded parts of the curved pipes, such as elbows. However, there have been no cases of the application of induction bending to the piping of nuclear power plants. In this study, the applicability of the P91 induction bending piping for the sodium-cooled fast reactor PGSFR was validated through high temperature low cycle fatigue tests and creep tests using P91 induction bending pipe specimens. The tests confirmed that the materials sufficiently satisfied the fatigue life and the creep rupture life requirements for P91 steel at 550 ℃ in the ASME B&PV Code, Sec. III, Div. 5. The results show that the effects of heating and bending by the induction bending process on the material properties were not significant and the induction bending process could be applicable to piping system of PGSFR well.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.359-374
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• 2024

A high-fidelity numerical analysis methodology was proposed for evaluating the fuel rod cladding integrity of a Prototype Gen IV Sodium Fast Reactor (PGSFR) during normal operation and Design basis events (DBEs). The MARS-LMR code, system transient safety analysis code, was applied to analyze the DBEs. The results of the MARS-LMR code were used as boundary condition for a 3D computational fluid dynamics (CFD) analysis. The peak temperatures considering HCFs satisfied the cladding temperature limit. The temperature and pressure distributions were calculated by ANSYS CFX code, and applied to structural analysis. Structural analysis was performed using ANSYS Mechanical code. The seismic reactivity insertion SSE accident among DBEs had the highest peak cladding temperature and the maximum stress, as the value of 87 MPa. The fuel cladding had over 40 % safety margin, and the strain was below the strain limit. Deformation behavior was elucidated for providing relative coordinate data on each active fuel rod center. Bending deformation resulted in a flower shape, and bowing bundle did not interact with the duct of fuel assemblies. Fuel rod maximum expansion was generated with highest stress. Therefore, it was concluded that the fuel rod cladding of the PGSFR has sufficient structural safety margin during DBEs.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1081-1090
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• 2024

The liquid Lead-bismuth eutectic is used as the coolant for Gen-IV reactor concepts. However, due to its strong corrosive and high operating temperature, it is difficult to accurately measure the flow rate in long-term operating conditions. Venturi flowmeter is a simple structured flowmeter, which plays a very important role in the flow measurement of high-temperature liquid metals, especially since the existing flowmeters are difficult to be competent. It has the advantages of easy maintenance and stable operation. Therefore, it is necessary to study the operating conditions of the venturi flowmeter under high-temperature conditions. This work performs a series of simulations of the fluid-solid interaction between the flow liquid metal and venturi flowmeter with COMSOL software, including the dimensional sensitivity analysis of the venturi flowmeter to explore the most suitable structure and parameters for liquid heavy metal, the sensitivity analysis of the geometric parameters of the venturi tube on the varying conditions. It shows that when the contraction angle of the venturi flowmeter is 33°, the diffusion angle is 13°, the diameter of the throat is 8 mm, and the temperature of the lead-bismuth eutectic is 733.15 K, it is most suitable for the measurement in the lead-bismuth circuit.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1125-1134
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• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.257-264
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• 2015

To maintain sustainability of nuclear energy as an important energy source, both safety problem and Spent Nuclear Fuels(SNFs) problem should be solved. In case of Gen-IV reactors such as fast reactor, SNFs can be used as fuels by using fast neutrons. It can be a suitable treatment method of high-level waste in near future. Liquid metals such as Sodium or Lead-Bismuth Eutectic (LBE) can be possibly used as a coolant to use fast neutrons. In this paper, it was described that natural circulation parameter studies, design analyses, material selections and a completion of facilities. To develop a natural circulation facility, thermal hydraulic analyses were performed. Installation technique of liquid metal natural circulation were secured.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.11
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• pp.1257-1264
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• 2014

Efforts are presently underway for developing an optimal design methodology for GEN-IV nuclear reactors based on target failure probabilities. A typical example is the system-based code, in which the results are represented in the form of partial safety factors (PSFs). Thus, a PSF is one of the crucial elements in either component design or integrity assessment based on target failure probabilities during the operation period. In the present study, a procedure for calculating the PSF of a circumferential through-wall cracked pipe based on the elastic-plastic crack initiation criterion is established, in which the importance of each input variable is assessed. Elastic-plastic J-integrals are calculated using the GE/EPRI and reference stress methods, and the PSF values are calculated using both first- and second-order reliability methods. Moreover, the effect of statistical distributions of assessment variables on the PSF is also evaluated.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1611-1625
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• 2020

We propose a scaled-down experimental model of vertical air-natural convection channels by applying the modified Ishii-Kataoka scaling method with the assistance of numerical analyses to the Reactor Vault Cooling System (RVCS) of the Proto-type Gen-IV Sodium-cooled fast reactor (PGSFR) being developed in Korea. Two major non-dimensional numbers (modified Richardson and Friction number) from the momentum equation and Stanton number from the energy balance equation were identified to design the scaled-down experimental model to assimilate thermal-hydraulic behaviors of the natural convective air-cooling channel of RVCS. The ratios of the design parameters in the PGSFR RVCS between the prototype and the scaled-down model were determined by setting Richardson and Stanton number to be unity. The friction number which cannot be determined by the Ishii-Kataoka method was estimated by numerical analyses using the MARS-KS system code. The numerical analyses showed that the friction number with the form loss coefficient of 2.0 in the scale-down model would result in an acceptable prediction of the thermal-hydraulic behavior in RVCS. We also performed experimental benchmarking using the scaled-down model with the MARS-KS simulations to verify the appropriateness of the scale-down model, which demonstrated that the temperature rises and the average air flow velocity measured in the scale-down model.

• Food Science of Animal Resources
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• v.37 no.3
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• pp.464-468
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• 2017

The identification of pork in commercially processed meats is one of the most crucial issues in the food industry because of religious food ethics, medical purposes, and intentional adulteration to decrease production cost. This study therefore aimed to develop a method for the detection of pork adulteration in meat products using primers specific for pig mitochondrial DNA. Mitochondrial DNA sequences for pig, cattle, chicken, and sheep were obtained from GenBank and aligned. The 294-bp mitochondrial DNA D-loop region was selected as the pig target DNA sequence and appropriate primers were designed using the MUSCLE program. To evaluate primer sensitivity, pork-beef-chicken mixtures were prepared as follows: i) 0% pork-50% beef-50% chicken, ii) 1% pork-49.5% beef-49.5% chicken, iii) 2% pork-49% beef-49% chicken, iv) 5% pork-47.5% beef-47.5% chicken, v) 10% pork-45% beef-45% chicken, and vi) 100% pork-0% beef-0% chicken. In addition, a total of 35 commercially packaged products, including patties, nuggets, meatballs, and sausages containing processed chicken, beef, or a mixture of various meats, were purchased from commercial markets. The primers developed in our study were able to detect as little as 1% pork in the heat treated pork-beef-chicken mixtures. Of the 35 processed products, three samples were pork positive despite being labeled as beef or chicken only or as a beef-chicken mix. These results indicate that the developed primers could be used to detect pork adulteration in various processed meat products for application in safeguarding religious food ethics, detecting allergens, and preventing food adulteration.