• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.707-712
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• 1995

Grain boundary diffusion plays significant role in the fission gas release, which is one of the crucial processes dominating nuclear fuel performance. Gaseous fission products such as Xe and Kr generated inside fuel pellet have to diffuse in the lattice and in the grain boundary before they reach open space in the fuel rod. In the mean time, the grains in the fuel pellet grow and shrink according to grain growth kinetics, especially at elevated temperature at which nuclear reactors are operating. Thus the boundary movement ascribed to the grain growth greatly influences the fission gas release rate by lengthening or shortening the lattice diffusion distance, which is the rate limiting step. Sweeping fission gases by the moving boundary contributes to the increment of the fission gas release as well. Lattice and grain boundary diffusion processes in the fission gas release can be studied by 'tracer diffusion' technique, by which grain boundary diffusion can be estimated and used directly for low burn-up fission gas release analysis. However, even for tracer diffusion analysis, taking both the intragranular grain growth and the diffusion processes simultaneously into consideration is not easy. Only a few models accounting for the both processes are available and mostly handle them numerically. Numerical solutions are limited in the practical use. Here in this paper, an approximate analytical solution of the lattice and stationary grain boundary diffusion in a polycrystalline solid is developed for the tracer diffusion techniques. This short closed-form solution is compared to available exact and numerical solutions and turns out to be acceptably accurate. It can be applied to the theoretical modeling and the experimental analysis, especially PIE (post irradiation examination), of low burn up fission. gas release.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.248-254
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• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Journal of the Korean Society of Radiology
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• v.11 no.7
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• pp.571-578
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• 2017

Bubbles are generated by the boiling of the cooling water when an accident occurs in the reactor and then in order to measure the void fraction, the Optical Fiber Probe(OFP) and optical camera are used in thermal hydraulic safety research. However, such an optical method is not suitable for measuring the void fraction in a $17{\times}17$ array of fuel rods due to the geometrical limitations. This study was conducted as a preliminary study using x-ray system and various phantoms before applying to rod bundles. Through radiographic and tomographic experiments, the tube voltage of the x-ray generator was 130 kVp and the tube current was 1 mA. In addition, it is possible to measure the hole of 1mm in size visually through the bubble resolution phantom, and it is confirmed that the contrast is relatively decreased in the inside of the freon in the case of the contrast evaluation using the road phantom. However, we could obtain good image without distortion when reconstructing the image. Bubble generation phantom experiments were used to confirm the flow direction of the bubbles and to acquire tomography images. The image J tool was used to measure the void fraction of 18 % for a single tomography image. This study has carried out previous researches for the measurement of the bubble rate around the nuclear fuel and could be used as a basic research for continuous research.

• 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.16 no.2
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• pp.106-115
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• 1984

Core safety limits define reactor operating conditions and parameters that will assure fuel rod and reactor system's integrity. Limiting safety system settings (LSSS) programmed into reactor protection system (RPS) then ensure a rapid reactor trip to prevent or suppress conditions which might violate the core safety limits. Generation of the LSSS must properly take into account uncertainties in both calculated and measured parameters in order to assure, with an appropriate degree of confidence, that the RPS will protect the core safety limits. Reviewed in this report are Westinghouse RPS setpoint generation philosophy, methodology of safety limit development and LSSS generation procedure. The Westinghouse RPS trip setpoint generation methodology has been established based on the calculation of core safety limits and the selection of LSSS allowing appropriate uncertainties in a conservative manner. Such conservative values of setpoint assure a high degree of core protection against fuel melting and occurrence of DNB.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2004.11a
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• pp.37-40
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• 2004

The impact/sliding wear tests have been performed in high temperature high pressure water in order to evaluate the effect of spring shape on the wear behavior of a spring supported tube for nuclear fuel fretting study. The results indicate that the tube wear volume and the size of the wear scar are closely related to each spring shape. From the analysis of the wear scar, it is possible to extract the real worn area (Aw) from the size of the wear scar (At). In addition, we found that the wear volume has a linear relation with the real worm area rather than the size of wear scar and this was only determined by each spring shape in the high temperature and pressure water condition. From the above results, it is possible to evaluate the wear resistant spring using the correlation between the variation of the real worn area and the wear behavior at each spring.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.1373-1376
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• 2008

A supporting structure for a long tube bundle of a large diameter is considered in this paper. The primary purpose of the present study is to develop a spacer grid structure for a so-called "dual cooled nuclear fuel", which has been being studied for a nuclear power uprate. The outer diameter of the fuel rod increases considerably from the conventional one. So a completely new shape of the supporting structure (spacer grid) needs to be developed. One of the challenges is to insert a supporting tube into the cross points of the grid straps. To meet a supporting performance, the load vs. displacement characteristics should be obtained. So the present study focuses on the finite element analysis technology to evaluate the characteristics through a parametric study. As a result, major influencing parameters are investigated for an optimized spacer grid design.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.14-18
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• 2020

Korea Atomic Energy Research Institute (KAERI) has carried out various neutronics experiments in the commissioning stage of the Jordan Research and Training Reactor (JRTR), and this paper introduces the results of first criticality prediction and experiment for the JRTR. The Monte Carlo Code for Advanced Reactor Design and analysis (McCARD) with the ENDF/B-VII.0 nuclear library was used for prediction calculations in the process of the first criticality approach, which was performed to provide reference for the first criticality experiment. In the experiment, fuel loading was carried out by measuring the inverse multiplication factor (1/M) to predict the number of fuel assemblies at the first criticality, and the first critical was reached on April 25, 2016. Comparing the first criticality prediction and experiment, the calculated and measured CAR (Control Absorber Rod) heights for the first criticality were 575 mm and 570.5 mm, respectively, that is, the difference between the two results was approximately 5 mm. From this result, it was confirmed that JRTR manufacturing and various experiments had successfully progressed as designed.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1181-1190
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• 2009

Knowing more about the Local Power Density (LPD) at the hottest part of a nuclear reactor core can provide more important information than knowledge of the LPD at any other position. The LPD at the hottest part needs to be estimated accurately in order to prevent the fuel rod from melting in a nuclear reactor. Support Vector Machines (SVMs) have successfully been applied in classification and regression problems. Therefore, in this paper, the power peaking factor, which is defined as the highest LPD to the average power density in a reactor core, was estimated by SVMs which use numerous measured signals of the reactor coolant system. The SVM models were developed by using a training data set and validated by an independent test data set. The SVM models' uncertainty was analyzed by using 100 sampled training data sets and verification data sets. The prediction intervals were very small, which means that the predicted values were very accurate. The predicted values were then applied to the first fuel cycle of the Yonggwang Nuclear Power Plant Unit 3. The root mean squared error was approximately 0.15%, which is accurate enough for use in LPD monitoring and for core protection that uses LPD estimation.