• Journal of the Korean Society for Nondestructive Testing
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• v.34 no.3
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• pp.254-259
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• 2014

A pressurized heavy water reactor (PHWR) core has 380 fuel channels contained and supported by a horizontal cylindrical vessel known as the calandria, whereas a pressurized water reactor (PWR) has only a single reactor vessel. The pressure tube, which is a pressure-retaining component, has a 103.4 mm inside diameter ${\times}$ 4.19 mm wall thickness, and is 6.36 m long, made of a zirconium alloy (Zr-2.5 wt% Nb). This provides support for the fuel while transporting the $D_2O$ heat-transfer fluid. The simple tubular geometry invites highly automated inspection, and good approach for all inspection. Similar to all nuclear heat-transfer pressure boundaries, the PHWR pressure tube requires a rigorous, periodic inspection to assess the reactor integrity in accordance with the Korea Nuclear Safety Committee law. Volumetric-based nondestructive evaluation (NDE) techniques utilizing ultrasonic and eddy current testing have been adopted for use in the periodic inspection of the fuel channel. The eddy current testing, as a supplemental NDE method to ultrasonic testing, is used to confirm the flaws primarily detected through ultrasonic testing, however, eddy current testing offers a significant advantage in that its ability to detect surface flaws is superior to that of ultrasonic testing. In this paper, effectiveness of flaw detection and the depth sizing capability by eddy current testing for the inside surface of a pressure tube, will be introduced. As a result of this examination, the ET technique is found to be useful only as a detection technique for defects because it can detect fine defects on the surface with high resolution. However, the ET technique is not recommended for use as a depth sizing method because it has a large degree of error for depth sizing.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1526-1531
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• 2024

¹³¹I is a fission product produced in a nuclear reactor by irradiating tellurium dioxide, with a half-life of 8.02 day. The most important and widely used method for making ¹³¹I is irradiation using a nuclear reactor and post-irradiation followed by dry distillation. The advantage of the dry distillation process is that the process and the equipment are relatively simple, namely TeO₂ (m.p. 750 ℃), which can withstand heating during reactor irradiation. Based on TeO₂ irradiation by neutron following the technique of dry distillation was explained for production of ¹³¹I on a large scale. A dry distillation followed the radioisotope production operation using the 30 MW GA Siwabessy nuclear reactor to meet national demand. TeO₂ targets are 25 and 50 g irradiated for 87-100 h. The resulting ¹³¹I activity is 20.29339-368.50335GBq. According to the requirements imposed on the radionuclide purity of the preparation, the contribution of ¹³¹I training in the resulting preparation was not less than 99.9 %

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.291-303
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• 2020

The Kori-Unit 1 nuclear power plant, which is scheduled to be decommissioned after permanent shutdown, is expected to generate large amounts of various types of radioactive waste during the decommissioning process. Among these, nuclear reactors and internal structures have high levels of radioactivity and the dismantled structure must have the proper size and weight on the primary side. During decommissioning, it is important to prepare an appropriate and efficient disposal method through analysis of the disposal status and the legal restrictions on wastes generated from the reactors and internal structures. Nuclear reactors and internal structures generate radioactive wastes of various levels, such as medium, very low, and clearance. A radiation evaluation indicates that wastes in the clearance level are generated in the reactor head and upper head insulation. In this study, a clearance waste safety evaluation was conducted using the RESRAD-RECYCLE code, which is a safety evaluation code, based on the activation evaluation results for the clearance level wastes. The clearance scenario for the target radioactive waste was selected and the maximum individual and collective exposure doses at the time of clearance were calculated to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. The evaluation results indicated that the doses were significantly low, and the clearance criteria were satisfied. Based on the safety assessment results, an appropriate metal recycle and disposal method were suggested for clearance, which are the subject of the deregulation of internal structures of nuclear power plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.333-349
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• 2013

As the capacity of spent nuclear fuel storage pool at reactor sites becomes saturated in ten years, long term dry storage strategy has been recently discussed as an alternative option in Korea. In this study, we reviewed safety-criteria-related research results on spent nuclear fuel performance and integrity under long-term dry storage and proposed the direction and the scope of future domestic research and development. Creep and hydride effect in relation to the embrittlement are known to be the major degradation mechanisms of the spent fuels during the long term dry storage. However, recent research results showed that hydride reorientation and hydride embrittlement are one of the most critical factors to the spent fuel integrity. Accordingly safety criteria of US and Japan for the storage system are basically founded on those mechanisms. However, in Korea, not only in-pile but out-of-pile experimental data have not been generated to understand fuel cladding degradation and to determine the criteria to ensure the safety. In addition, the transient behavior of the spent fuel during transportation also needs to be thoroughly examined. Therefore, various experimental research and development will be required to establish our own safety criteria for future long-term dry storage of domestic spent fuels.

• Fire Science and Engineering
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• v.26 no.6
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• pp.38-44
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• 2012

The purpose of fire protection for the nuclear power plants (NPPs) is to ensure safe shutdown state of the reactor, to minimize the release of radioactive materials to the environment, to provide physical safety of the on-site personnel, and to limit the property damage. Fire protection and extinguishing equipments are one of the important protection measures based on the defense-in-depth concept, which can promptly detect and control and extinguish those fires that do occur, thereby limiting fire damage. However, a separate evaluation process might be additionally necessary for the construction permit and operating license because the fire protection laws of the NEMA for installation standards of the fire protection systems is not fully characterized for the NPPs. It is also not easy to implement the regulations such as the performance based design concept for fire protection system of the NPPs which are characterized for a relatively low density of employee. This study suggests a guideline for the improvement of the technical standards for fire protection systems of the NPPs by evaluating the fundamental problems drawn by reviewing laws and regulatory guides relevant to fire protection and by evaluating the applicability of the KEPIC FPN in domestic nuclear power plants.

• 제어로봇시스템학회:학술대회논문집
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• 2004.08a
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• pp.1516-1519
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• 2004

A tele-operated robot should be used to maintain and inspect nuclear power plants to reduce the radiation exposure to the human operators. During an overhaul of the nuclear power plants in Korea, a ROV(Remotely Operated Vehicle) may enter a cold-leg connected to the reactor to examine the state of the thermal sleeve and it's position in the safety injection nozzle. To measure the positions of the thermal sleeve or scratches from the video images captured during the examination, the camera parameters should be identified. However, the focal length of the CCD camera could be increased to a close up of the target and the aspect ratio and the center of the image could also be varied with capturing devices. So, it is desired to self-calibrated the intrinsic parameters of the camera and capturing device with the video images captured during the examination. In the video image of the safety injection nozzle, two or more circular grooves around the nozzle are shown as ellipse contours. In this paper, we propose a camera self-calibration method using a single image containing two circular grooves which are the greatest circles of the cylindrical nozzle whose radius and distance are known.

• Corrosion Science and Technology
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• v.13 no.5
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• pp.178-185
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• 2014

Alloy 617 is considered as a candidate Ni-based superalloy for the intermediate heat exchanger (IHX) of a very high-temperature gas reactor (VHTR) because of its good creep strength and corrosion resistance at high temperatures. Helium is used as a coolant in a VHTR owing to its high thermal conductivity, inertness, and low neutron absorption. However, helium inevitably includes impurities that create an imbalance in the surface reactivity at the interface of the coolant and the exposed materials. As the Alloy 617 has been exposed to high temperatures at $950^{\circ}C$ in the impure helium environment of a VHTR, the degradation of material is accelerated and mechanical properties decreased. The high-temperature strength, creep, and corrosion properties of the structural material for an IHX are highly important to maintain the integrity in a harsh environment for a 60 year period. Therefore, an alloy superior to alloy 617 should be developed. In this study, the mechanical and high-temperature corrosion properties for Ni-Cr alloys fabricated in the laboratory were evaluated as a function of the grain boundary strengthening and alloying elements. The ductility increased and decreased by increasing the amount of Mo and Cr, respectively. Surface oxide was detached during the corrosion test, when Al was not added to alloy. However the alloy with Al showed improved oxide adhesive property without significant degradation and mechanical property. Aluminum seems to act as an anti-corrosive role in the Ni-based alloy.

• Journal of the Korean Society of Safety
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• v.14 no.3
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• pp.93-100
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• 1999

Decomposition efficiency, power consumption, and applied voltage of CFC(Chlorofluorocatbon) were investigated by SPCP(surface induced discharge plasma chemical processing) reactor to obtain optimum process variables and maximum decomposition efficiencies. Decomposition efficiency of CFC-12 with various electric frequencies(5~50kHz). flow rates (100~1,000mL/min), initial concentrations(100~1,000ppm), electrode materials(W, Cu, Al). electrode thickness(1, 2, 3mm) and reference gases($N_2$, $O_2$, air) were measured and the products were analyzed with FT-IR. Experimental results showed that at the frequency of 10kHz, the highest decomposition efficiency of 92.7% for CFC-12 were observed at the power consumptions of 29.6W. respectively, and that decomposition efficiency decreased with increasing frequency above 20kHz and decomposition efficiency per unit power were 3.13%/W for CFC-12. Decomposition efficiency was increased with increasing residence times and with decreasing initial concentration of pollutants. Decomposition efficiency was increased with increasing thickness of discharge electrode and the highest decomposition efficiency was obtained for the electrode diameter of 3m. As the electrode material, decomposition efficiency was in order that tungsten(W), copper(Cu), aluminum (Al). Decomposition of CFC-12 in the reference gas of $N_2$ showed the highest efficiency among three reference gases, and then the effect of reference gas on the decomposition efficiency decreased in order of air and $O_2$. The optimum power for the maximum decomposition efficiency was 25.3W for CFC.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.38 no.1
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• pp.71-79
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• 2014

A parallelization of the bi-conjugate gradient solver for the pressure equation of the CUPID (component unstructured program for interfacial dynamics) code, which was developed for analyzing the components of a pressurized water-cooled reactor, was studied in a symmetric multi-processing system. The parallel performance was investigated for three typical parallel programming models (MPI, OpenMP, Hybrid) by solving incompressible backward-facing step flow at various grid resolutions. It was confirmed that parallel performance was low when problem size was small or the memory requirement for each thread was considerably higher than the cache memory. Furthermore, it was shown that MPI was better than OpenMP regardless of the problem size, and Hybrid was the best when the number of threads was relatively small.

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