• Nuclear Engineering and Technology
• /
• v.40 no.4
• /
• pp.255-268
• /
• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.

• Journal of the Korean Society of Radiology
• /
• v.11 no.6
• /
• pp.531-538
• /
• 2017

Agricultural products produced in the agricultural area around the nuclear power plant are radioactive contamination, which can cause radioactive contamination to the human body. The purpose of this study was to investigate the limit of the radioactivity concentration $^{90}Sr$ for the internal exposure dose evaluation by ingesting the agricultural products collected around the nuclear power plant. The results of the gamma-isotope element analysis were freshly <0.0166-0.0336 Bq / kg for all samples and for artificial radionuclides not detected, and fresh <0.00586-0.0421 Bq / kg for Chinese cabbage, The freshness was 0.106 Bq / kg, and the freshness was 0.0114-0.0901 Bq / kg. 0.0177%, 0.0222%, 0.0376% and 0.00243%, respectively, for Chinese cabbages and large roots, which is lower than the legal standard value of $1mSv/yr{\cdot}man%$. It is considered that the formulas need to be broadly evaluated for the foods consumed by children and adults, taking into consideration the age of the food and the diet

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.1 no.1
• /
• pp.73-79
• /
• 2003

In a pressurized water reactor, radioactive corrosion products (CRUD) in primary coolant system are one of the major sources for the occupational radiation exposure of the personnel in a nuclear power plant. Through the recent trend of long term fuel cycle in a nuclear power Plant, radioactive corrosion products deposited in reactor core for a long time are also the cause of Axial Offset Anomaly (AOA) having m effect on reactor power by the hideout of boron. CRUD consist primarily of magnetite, nickel ferrite, cobalt ferrite, and so on. They have the characteristic of strong magnetism. Therefore it is performed the conceptual design to develop the filter which removes the CRUD by magnetic field that is generated by an arrangement of permanent and electric magnets. Contrary to the conventional filter, the proposed filter does not interrupt the fluid flow, so there is little pressure drop and it can be used continuously. It is expected to be applied as one of the technologies for removal of the CRUB.

• Korean Journal of Materials Research
• /
• v.27 no.10
• /
• pp.552-556
• /
• 2017

Evaluation of the durability and stability of materials used in power plants is of great importance because parts or components for turbines, heat exchangers and compressors are often exposed to extreme environments such as high temperature and pressure. In this work, high-temperature corrosion behavior of 316 L stainless steel in a carbon dioxide environment was studied to examine the applicability of a material for a supercritical carbon dioxide Brayton cycle as the next generation power plant system. The specimens were exposed in a high-purity carbon dioxide environment at temperatures ranging from 500 to $800^{\circ}C$ during 1000 hours. The features of the corroded products were examined by optical microscope and scanning electron microscope, and the chemical compound was determined by x-ray photoelectron spectroscopy. The results show that while the 316 L stainless steel had good corrosion resistance in the range of $500-700^{\circ}C$ in the carbon dioxide environment, the corrosion resistance at $800^{\circ}C$ was very poor due to chipping the corroded products off, which resulted in a considerable loss in weight.

• Proceedings of the KIEE Conference
• /
• 2011.07a
• /
• pp.1191-1192
• /
• 2011

We released new triple redundant digital excitation system with real time generator-turbine simulator. One of its great merits is the real time generator-turbine simulator when it was compared with the other products. If excitation system is tripped by unexpected faults, Maintenance man can do easily performance test of digital excitation control board, sequence relay and thyristor switching device of phase controlled rectifier without manufacturer's support. For the verification of this system, It was tested with an actual excitation system implemented on 5kVA M-G Set. After finishing the tests, the trial product will be installed and operated at a 500MW thermal power plant.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.02a
• /
• pp.64-72
• /
• 2004

The dryer (ten per unit) are operating to remove tritium in PHWR(Pressurized Heavy Water Reactor). There are coming out heavy water adsorbent from operating the dryer (95 drums for ten year per unit) The amount of radioactivity of heavy water adsorbent almost exceed ninety times more than disposal limit-in-itself showed by The Ministry of Science and Technology. It has to be disposed whole radioactive waste products, however there are problems of increase at the expense of their permanent disposal. In this research, We have studied how to remove kinds of nuclear materials and amount of tritium with in heavy water adsorbent. As the result we could develop disposal equipment and apply it. D20 adsorbent have to contain below Gamma nuclide O.3Bq/g and tritium 100Bq/g "The Regulation for disposal of the radioactivity wastes" showed by The Ministry of Science and Technology. There fore. So as to remove amount of tritium and kinds of nuclear materials (DTO) we needed a equipment. Also we have studied how to remove effectively radioactivity with in Adsorbent. As cleaning heavy water adsorbent and drying on each condition (temperature for drying and hours for cleaning). Because there is something to return heavy water adsorbent by removing impurities within adsorbent when it is dried o high temperature. After operating, we have been applying this research to the way to dispose heavy water adsorbent. Through this we could reduce solid waste products and the expense of permanent disposal of radioactive waste products and also we could contribute nuclear power plant run safely. According to the result we could keep the best condition of radiation safety super vision and we could help people believe in safety with Radioactivity wastes control for harmony with Environment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.1
• /
• pp.31-40
• /
• 2005

High temperature - high pressure apparatus was developed to simulate nickel fewite corrosion products which were main compositions of the radioactive crud in the nuclear power plant. Corrosion product similar to the crud was obtained by a tube accumulator system. Nickel alloy (Inconel 690) and carbon steel (SA106 Gr. C) were corroded at 270 $\^{circ}C$ in the corrosion product generator. Ni ions and Fe ions dissolved by corrosion reaction were able to be transported to the accumulator because the crud generation mechanism was the solubility change with temperature. To evaluate the properties of simulated corrosion products, scanning electron microscope (SEM) observation and EDAX analysis were performed. SEM observation of corrosion product showed the needlelike or crystal structure of oxide depending on precipitating location. The crystal oxide was the nickel ferrite, which was similar to the crud in nuclear power plants.

• Proceedings of the Korea Concrete Institute Conference
• /
• 2002.05a
• /
• pp.1015-1020
• /
• 2002

Recycling of coal combustion by-product(Ash) are becoming more improtant in the utilization business as a result of the increased use of NOx reduction technologies at coal-fired power plants. Current disposal methods of these by-products create not only a loss of profit for the power industry, but also environmental concerns that breed negative public opinion. This research made Precast block for environment-friendly secondary product and compare strength special quality of this block with existent common use brick and analyze application possibility in situ with a reserve experiment that measure strength property and manufacture method to handle coal ash produced in Bo-ryung thermoelectric power plant.

• Nuclear Engineering and Technology
• /
• v.42 no.1
• /
• pp.79-88
• /
• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• 한국연소학회:학술대회논문집
• /
• 2006.10a
• /
• pp.137-144
• /
• 2006

Oil-fired power plants usually use several burners and the combustion air is supplied to each burner through the complicated duct which is called windbox. A windbox should be designed to supply combustion air to each burner evenly but, due to the complicated duct shape, flow distribution in the windbox is unbalanced and uneven supplies of combustion air to each burner are induced by these unbalanced flow distribution in the windbox. These flow patterns tend to make flame unstable, increase the formation of pollutants and lower the overall combustion efficiency. To prevent these disadvantages, flow patterns in the windbox should be investigated for the uniform flow distribution. In this study, computational simulation method was used to investigate the flow distribution in the windbox and measured the velocities at the exit of burners in the real windbox to compare with CFD results. The results show two significant flow patterns. One is that the flow rates of each burner are different from each other and this means that all burners operate in different conditions of air to fuel ratio. The other is that the flow distribution at the exit of each burner is not axi-symmetric although the burner shape is axi-symmetric and this increases the pollutant products like CO.