• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1485-1494
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• 2023

The bow of the nuclear fuel assembly is a well-known phenomenon. One of the vital criteria during the history of nuclear fuel development has been fuel assembly's mechanical stability. Once present, the fuel assembly bow can lead to safety issues like excessive water gap and power redistribution or even incomplete rod insertion (IRI). The extensive bow can result in assembly handling and loading problems. This is why the fuel assembly's bow is one of the most often controlled geometrical factors during periodic fuel inspections for VVER when compared e.g. to on-site fuel rod gap measurements or other instrumental measurements performed on-site. Our proposed screening method uses existing video records for fuel inspection. We establish video frames normalization and aggregation for the purposes of bow measurement. The whole process is done by digital image processing algorithms which analyze rotations of video frames, extract angles whose source is the fuel set torsion, and reconstruct torsion schema. This approach provides results comparable to the commonly utilized method. We tested this new approach in real operation on 19 fuel assemblies with different campaign numbers and designs, where the average deviation from other methods was less than 2 % on average. Due to the fact, that the method has not yet been validated during full scale measurements of the fuel inspection, the preliminary results stand for that we recommend this method as a complementary part of standard bow measurement procedures to increase measurement robustness, lower time consumption and preserve or increase accuracy. After completed validation it is expected that the proposed method allows standalone fuel assembly bow measurements.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.267-279
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• 1983

An analytical system has been established for scrutinizing both uranium- and plutonium-fueled lattices moderated by light water. This system consists of two primary codes. One is a unit cell program called KICC, which has theoretical foundation on the models of GAM and THERMOS incorporated with appropriate approximate treatments for various phenomena, whereas the other is a multi-dimensional diffusion-depletion program entitled KIDD. The adequacy of this system is verified by performing extensive benchmark calculations on a variety of critical experiments. The average value of effective multiplication factors for the selected nineteen UO$_2$ critical experiments of heterogeneous lattice structure is calculated to be 1.0006 with a standard deviation of 0.0039. Power distributions have also been calculated for some critical experiments fueled with both uranium and plutonium of varying concentrations. The maximum percentage difference between the measured and calculated power distributions appears to be less than 5%. This result, together with the previously reported result, illustrates that the KICC/KIDD system is a very effective tool for the analysis of a light water reactor core.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.87-93
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• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.225-232
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• 2008

An estimation has been made for inventories and characteristics of spent nuclear fuel(SNF) to be generated from existing and planned nuclear power plants based on the 3rd Basic Plan for Electric Power Demand and Supply. The characteristics under consideration in this study are dimensions, a fuel rod array, a weight, $^{235}U$ enrichment, and the discharge burnup in terms of fuel assembly. These are essentially needed for designing a pyroprocessing facility. It is appeared that the anticipated quantity by the end of 2077 is about 23,000 tU for PWR spent nuclear fuel. It is revealed that the proportion of SNF with the initial $^{235}U$ enrichment below 4.5 weight percent(wt.%) is approximately 95 % in total. For SNF with 16$\times$16 fuel rod array the proportion is expected approximately 74% in total. It appears that the average burnup of SNF will be 55 GWd/tU after the medium and/or latter part of 2010s while the average burnup is 45 GWd/tU at present. Finally, a requirement in terms of reference SNF for designing the pyroprocessing facility has been derived from the above-mentioned results. The anticipated SNF seems to be 16$\times$16 Korean Standard Fuel Assembly with a cross section of 21.4 cm$\times$21.4 cm, a length of 453 cm, a mass of 672 kg, the initial $^{235}U$ enrichment of 4.5 wt.%, and the discharge burnup of 55 GWd/tU.

• Journal of Electrochemical Science and Technology
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• v.10 no.2
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• pp.104-114
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• 2019

Energy is an essential driving force for modern society. In particular, electricity has become the standard source of power for almost every aspect of life. Electric power runs lights, televisions, cell phones, laptops, etc. However, it has become apparent that the current methods of producing this most valuable commodity combustion of fossil fuels are of limited supply and has become detrimental for the Earth's environment. It is also self-evident, given the fact that these resources are non-renewable, that these sources of energy will eventually run out. One of the most promising alternatives to the burning of fossil fuel in the production of electric power is the proton exchange membrane (PEM) fuel cell. The PEM fuel cell is environmentally friendly and achieves much higher efficiencies than a combustion engine. Water management is an important issue of PEM fuel cell operation. Water is the product of the electrochemical reactions inside fuel cell. If liquid water accumulation becomes excessive in a fuel cell, water columns will clog the gas flow channel. This condition is referred to as flooding. A number of researchers have examined the water removal methods in order to improve the performance. In this paper, a new water removal method that investigates the use of vibro-acoustic methods is presented. Piezo-actuators are devices to generate the flexural wave and are attached at end of a cathode bipolar plate. The "flexural wave" is used to impart energy to resting droplets and thus cause movement of the droplets in the direction of the traveling wave.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.115-120
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• 2000

The analysis of flux distribution under stead-state in large power reactors with assymetry reactivity insertions requires the use of three-dimensional diffusion calculations. For the purpose, consistently formulated modern nodal methods based on higher order interface techniques have become popular tools for flux distributions in large commercial nuclear reactors. Among the earlier developments, the nodal Green's function method obtains its nodal interface equation from the transverse-integrated integral diffusion equation using a finite-medium Green's function. In this method, the outgoing current from a node surface is formulated as a response of the incoming currents and the spatially integrated neutron source within the same node. The well-known nodal expansion method is also based on an interface partial current formulation. Nodal methods high-level interface variables, i.e., interface net current and flux, may be more computationally efficient than the nodal Green's function method because they have one fewer unknown per interface. The Analytic Nodal Method(ANM), which can be classified as an interface net current technique and, was faster in solving some standard benchmark problems than the other two methods.(omitted)

• Korean Journal of Food Science and Technology
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• v.36 no.1
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• pp.183-187
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• 2004

Radioactivity in foodstuffs was surveyed for reference in amending regulation on the maximum permitted levels of radioactive contamination of foodstuffs. Most domestic and imported (?) foodstuffs were sampled, some domestic items collected around nuclear power plants to compare site-specific contamination. The collected samples were dried and ashed. Radioactivity in foodstuffs was measured using HPGe gamma spectrometer, Cs-137 activity ranged from 0.025-0.053, 0.045-0.500, 0.062-0.105, 0.025-1.151, 0.021-0.145 and 0.046-0.155 Bq/kg-fresh in cereals, pulses, mot vegetables (potato), ginsengs, meat, and marine products, respectively, with imported dried ginseng showing the highest radioactivity, Results reveal radioactivity in foodstuffs collected in 2002 is far below the maximum permitted levels of 370 Bq/kg. No significant differences were observed in radioactivity among sampling sites and between domestic and imported foodstuffs.

• Journal of the Korean Society of Radiology
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• v.14 no.7
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• pp.857-862
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• 2020

In March 2011, a tsunami off Japan caused radioactive material that had seeped into the sea from the Fukushima nuclear accident to flow to the Pacific Ocean, causing pollution to sea life. For a comparative evaluation with the area surrounding the site of a nuclear power plant by the release of radioactive materials, an area 20 to 30 km away from the emergency protection plan area was selected as a comparative point considering weather conditions, population distribution, etc. In addition, the government intends to analyze the minimum detection radiation received by residents around the nuclear power plant and evaluate the effective dose. Analysis of tritium radiation from water samples showed that most of the samples were not detected and that 0.0014 % to 0.777 % of the annual legal standard of 1 mSv for the general public had little effect on the human body. Therefore, the measurement and analysis of water samples around the nuclear power plant site is expected to help relieve anxiety, such as exposure to the general public and neighboring residents due to radiation release.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.29 no.9 s.240
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• pp.1253-1261
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• 2005

Cast stainless steel (CSS) is thermally aged by a long term exposure in the range of nuclear power plant operating temperature. The thermal aging is a cause of concern for the continued safe and reliable operation of CSS nuclear components. Therefore, an assessment of degradation in material properties of these components has been importantly considered. In this study the ball indentation tests were performed on four cast stainless steels aged at $400^{\circ}C$ for 3600 hours, to investigate the applicability of ball indentation test to the assessment of aging degradation of cast stainless steels. Thus, the reliability of ball indentation test for aged CSS was analyzed by evaluating the scattering of data tested from each material and by comparing tensile properties obtained from ball indentation test and standard tensile test. Also, the tensile properties of aged CSS obtained from ball indentation test were compared with those predicted by the evaluation procedure developed on the basis of material database for aged CSS.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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• pp.823-832
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• 2003

To define the causes of cladding degradation which can take place during the operation of nuclear power plants, it is required to develop the new fracture toughness test of spent fuel cladding. The fracture toughness of Zircaloy-4 cladding was estimated using the recently developed KAERI embedded Charpy (KEC) specimen. Axially notched KEC specimens cut directly from unirradiated fuel claddings, were tested in a way similar to the standard toughness test method of a Single Edge Bending (SEB) specimen. The results of KEC fracture toughness test at room temperatures were discussed and compared with those of the previous other studies. In conclusions, even though the KEC fracture toughness test of nuclear fuel claddings was easier and more reliable than those developed earlier, the results from the cladding fracture tests were not the material characteristics but the specific fracture parameters which were deeply related to the specification of claddings. In addition, the phenomenon of a thickness yielding was not observed from the fracture surface. It was closely related to the fact that the plane strain condition of the KEC specimen was changed to the plane stress condition during crack advancing. It was also supported by the fractographic evidence that the formation of ductile dimples at the crack initiation became the similar appearance such as a quasi-cleavage after the sufficient crack advancing.