The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

In this study, we designed and manufactured a large angular contact ball bearing (LACBB) with low deformation using JIS-SUJ2 steel and analyzed changes in its structural characteristics and chemical composition upon heat treatment. The bearing was produced by hot forging and heat treatment including a quenching and tempering (Q/T) process, and its properties were analyzed using 4 mm thick specimens. A difference in the size distribution of the carbide in the outer and inner parts of the bearing was observed and it was confirmed that large and non-uniform carbide was distributed in the inner part of the bearing. After heat treatment, the hardness value of the outer part increased from 13.4 HRC to 61 HRC and the inner part increased from 8.0 HRC to 59.7 HRC. As a result of X-ray diffraction (XRD) measurements, the volume fraction of the retained austenite contained in the outer part was calculated to be 3.5~4.8 % and the inner part was calculated to be 3.6~5.0 %. The surface chemical composition and the content of chemical bonds were quantified through X-ray photoelectron spectroscopy (XPS), and a decrease in C=C bonds and an increase in Fe-C bonds were confirmed.

The Wolsong Nuclear Power Plant (NPP) operates an on-site spent fuel dry storage facility using concrete silo and vertical module systems. This facility must be safely maintained until the spent nuclear fuel (SNF) is transferred to an external interim or final disposal facility, aligning with national policies on spent nuclear fuel management. The concrete silo system, operational since 1992, requires an aging management review for its long-term operation and potential license renewal. This involves comparing aging management programs of different dry storage systems against the U.S. NRC’s guidelines for license renewal of spent nuclear fuel dry storage facilities and the U.S. DOE’s program for long-term storage. Based on this comparison, a specific aging management program for the silo system was developed. Furthermore, the facility’s current practices—periodic checks of surface dose rate, contamination, weld integrity, leakage, surface and groundwater, cumulative dose, and concrete structure—were evaluated for their suitability in managing the silo system’s aging. Based on this review, several improvements were proposed.

In the 3rd revision of NUREG-0800, which was revised in 2007, the calculation method for decay heat in the design of the Ultimate Heat Sink (UHS) for a pressurized water reactor is recommended to be based on the ANSI/ANS-5.1 method. This method employs a more complex decay heat calculation formula compared to the one introduced in Branch Technical Position ASB 9-2, which was presented in the 2nd revision. While most of the variables for decay heat calculation in ANSI/ANS-5.1 can be inferred from the methods outlined in the appendices, determining the fractions of fission products is not straightforward despite their significant impact on the results. When reviewing documents that evaluate decay heat using the ANSI/ANS-5.1 method, it is observed that they often adopt a conservative approach by assuming that the fraction of the most influential fission product is 100%. In this study, the fractions of each fission product presented in LLNL’s 2016 report were used to calculate decay heat, and the results were compared with the ASB 9-2 method and ORIGEN code results. The comparison showed that ANS 5.1 tends to yield higher decay heat values than ANS 9-2, particularly at the reference time of 1M seconds, while ORIGEN-ARP generally produced lower values. Therefore, it is concluded that even when using the ANSI/ANS-5.1 method with the fractions of each fission product for decay heat calculations in spent nuclear fuel wet or dry storage facility assessments, it provides a sufficiently conservative thermal evaluation.

After the Fukushima disaster, overseas nuclear power plants have established conditions for issuing a red alert in the event of fuel damage within the spent fuel pool and they have already implemented conditions for issuing a blue alert when fuel is exposed above the water surface. In South Korean nuclear power plants, a real-time monitoring system is in place to oversee the exposure of spent fuel to the surface within the spent fuel pool. To achieve this, a water level indicator gauge is installed within the spent fuel pool, allowing for continuous real-time monitoring. This paper conducted a comparative assessment of radiation levels from water level monitoring system in two units’ spent fuel pools based on the low water levels (1 feet from the storage rack), utilizing the radiation analysis code (MCNP).

After the decision of the Wolsong unit 1 permanent shutdown (2019), spent fuel stored in the spent fuel bay (hereafter, SFB) should be transported to a dry storage facility (MACSTOR or Canister) in order to decommission Wolsong unit 1. Accordingly, KHNP has established a shipment schedule for damaged fuel of Wolsong Unit 1 and is trying to complete the shipment according to the schedule. Wolsong is equipped with transportation casks and dry storage facilities, but baskets need to be manufactured separately. In addition, license approval is required for baskets, transport cask, and dry storage facilities for legal grounds to contain, transport, and store damaged fuels. In this paper, the initial model, upgrade model, and automation model of encapsulation equipment planned to be introduced in Canada to handle PHWR’s damaged fuel were compared, and the optimal model was selected in consideration of KHNP’s planned spent fuel shipment schedule. The PHWR’s damaged fuel encapsulation system is a system developed the PHWR’s damaged spent fuel to be handled in the same way as the existing PHWR when storing it in the dry storage facility and loading a basket for capsulation into transport cask. At the Gentilly-2 nuclear power plant in Canada, a manually operated encapsulation system was used due to the low quantity of damaged fuel, which can be encapsulated two bundles a day, and this model is an initial model. In the case of Wolsong Unit 1, it has about 300 damaged fuels, so it takes about nine months to work when using the initial model. The upgrade model developed to improve work efficiency and reliability has increased work efficiency through some automation, but it would take about eight months to process the damaged spent fuels of Wolsong Unit 1, and this model has not yet been manufactured and applied. Lastly, the automation model changed the work location outside the SFB and automated drainage/drying operations. It is easy to maintain and replace consumables because the work is carried out by lifting the damaged fuel to a shuttle outside the SFB surrounded by a shielding chimney. Considering the reduction of drainage/drying time, it is possible to save about four times as much time as the initial model. That is, if the automation model is used, it is judged that the supply of Wolsong Unit 1 can be processed in about two months. However, in terms of license, initial model and upgrade model are expected to be easier and the period is expected to be shortened. However, if licensing is carried out as soon as equipment design is completed, it is believed that the period can be shortened by parallel equipment manufacturing and licensing. It is judged that the best way to comply with the target schedule is to select an automation model with excellent work performance, develop equipment, and proceed with licensing at the same time. Accordingly, KHNP is in the process of designing equipment with the aim of using the automation model to take out damaged fuel for Wolsong Unit 1.

Since the September 11 terrorist attacks in the United States, concerns about intentional aircraft crashes into nationally critical facilities have soared in countries around the world. The United States government advised nuclear utilities to strengthen the security of nuclear power plants against aircraft crashes and stipulated aircraft crash assessment for new nuclear facilities. Interest in military missile attacks on nuclear facilities has grown after Russia attacked Ukraine’s Zaporizhzhia nuclear power plant, where spent nuclear power dry storage facility is operated. Spent nuclear fuel dry storage facilities in nuclear power plant sites should also strengthen security in preparation for such aircraft crashes. Most, but not all, spent nuclear fuel dry storage facilities in Europe, Japan and Canada are operated within buildings, while the United States and Korea operate dry storage facilities outdoors. Since all of Korea’s dry storage systems are concrete structures vulnerable to crash loads and are exposed to the outside, it is more necessary to prepare for aircraft crash terrorist attacks due to the Korea’s military situation. Residents near nuclear power plants are also demanding assessment and protective measures against such aircraft crashes. However, nuclear power plants, including spent nuclear fuel dry storage facilities, are strong structures and have very high security, so they are unlikely to be selected as targets of terrorism, and spent nuclear fuel dry storage systems are so small that aircraft cannot hit them accurately. Collected opinions on the assessment of aircraft crash accidents at spent nuclear fuel dry storage facilities in nuclear power plant sites were reviewed. In addition, the current laws and regulatory requirements related to strengthening the security of new and existing nuclear power plants against intentional aircraft crashes are summarized. Such strengthening of security can not only ensure the safety of on-site spent nuclear fuel dry storage facilities, but also contribute to the continuous operation of nuclear power plants by increasing resident acceptance.

More than 20,000 bundles of spent nuclear fuel are stored in the spent nuclear fuel storage pool of domestic nuclear power plants, and the dry storage facility project in the nuclear power plant site is being promoted as the saturation of the wet storage pool is imminent. Since bending or twisting of spent nuclear fuel is an important item in order to load spent nuclear fuel into a dry storage cask, PSE (Pool Side Examination) was performed to verify this. This paper describes whether it can be safely loaded into a dry storage cask based on the measurement results of bending or twisting of spent nuclear fuel. The nuclear fuel assembly is designed to prevent excessive assembly bending and twisting because it can cause interference during dry storage and handling due to factors such as differences in depletion of nuclear fuel rods, irradiation growth, and coolant flow during reactor operation. The bending of the nuclear fuel assembly is measured by establishing a Plumb Line to photograph the nuclear fuel assembly based on it, and calculating a pixel that images the distance between the support grid and the Plumb Line. The twisting of the nuclear fuel assembly is measured by forming a virtual vertical plane with two Plumb Lines, and based on this, the twisting angle of the lower fixed compared to the upper fixed. As a result of the measurement, the bending of spent nuclear fuel was about 0.0-10.2 mm, much lower than the reactor loading criteria of 15.0 mm, and in the case of twisting, about 0.0~2.2° much lower than the reactor loading criteria of 5.0°. Therefore, it was confirmed that spent nuclear fuel at domestic nuclear power plants was not affected by bending and twisting when loading into dry storage cask.

Korea Hydro & Nuclear Power (KHNP) is currently developing a vertical concrete dry storage module for the dry storage of used nuclear fuel within nuclear power plants. This module is designed with a structure consisting of cylinders, which can block the ingress of external air, thereby preventing Chloride-Induced Stress Corrosion Cracking (CISCC). However, due to the presence of these cylinder structures, unlike conventional dry storage systems, it cannot directly dissipate heat to the external atmosphere, making thermal evaluation an important issue. The SF dry storage module being developed by KHNP is a massive concrete structure of approximately 20 m × 10 m × 7 m in size, employing a vertical storage system. To demonstrate the safety of such a large structure, there is no alternative to conducting experiments with scaled-down models. Furthermore, according to NUREG-2215 Section 5.5.4, it is explicitly mentioned that design-verification testing can be performed using scaled-down models. In this paper, a 1/4 scaled-down model was constructed to perform thermal performance verification experiments, and the effectiveness of this model was analyzed using Computational Fluid Dynamics (CFD) methods. The analysis results indicated that there was not a significant difference in terms of maximum concrete temperature and air outlet temperature. However, a considerable difference was observed in the canister surface temperature. Therefore, it is concluded that careful consideration of natural convection heat transfer is necessary for the full application of the scaled-down model.

In Korea, borated stainless steel (BSS) is used as a storage rack in spent fuel pools (SFP) to maintain the nuclear criticality of spent fuels. As the number of nuclear power plants and the corresponding amount of spent fuels increased, the density in SFP storage rack also increased. In this regard, maintaining subcriticality of spent nuclear fuels became an issue and BSS was selected as the structural material and neutron absorber for high density storage rack. Since it is difficult to replace the storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and is specified in the ASTM A887-89 standard depending on the boron concentration from 304B (0.20-0.29% B) to 304B7 (1.75-2.25% B). Due to the low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr) 2B are formed as a secondary phase. Metallic borides could cause Cr depletion near it, which could decrease the corrosion resistance of the material. In this paper, the long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP conditions. Because the corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, the corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis is conducted using a scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, a hematite structure oxide film is formed, and pitting corrosion occurs on the surface of specimens. Most of the pitting corrosion is found at the substrate surface because the corrosion resistance of the substrate, which has low Cr content, is relatively low. Also, the oxidation reaction of B in the secondary phase has the lowest Gibbs free energy compared to other elements. Furthermore, oxidation of Cr has low Gibbs free energy, which means that oxidation of B and Cr could be faster than other elements. Thus, the long-term corrosion might affect the boron content and the neutron absorption ability of the material. Using boron’s high cross-section for neutrons, the neutron absorption performance of BSS was evaluated through neutron transmission tests. The effect of the corrosion behavior of BSS on its neutron absorption performance was investigated. Samples simulated to undergo up to 60 years of degradation before corrosion through accelerated corrosion testing did not show significant changes in the neutron shielding ability before and after corrosion. This can be explained in relation to the corrosion behavior of BSS. Boron was only leached out from the secondary phase exposed on the surface, and this oxidized secondary phase corresponds to about 0.17% of the volume of the total secondary phase. This can be seen as a very small proportion compared to the total boron content and is not expected to have a significant impact on neutron absorption performance.

The burnup of spent fuel is one of the important management items that must be managed before storing the fuel in dry storage facilities, as well as for transportation and disposal in the future. Currently, the burnup of spent fuel is managed by calculating the design burnup at the time of design and measuring the real burnup using in-reactor measurement devices. Furthermore, to ensure the reliability of such data, the burnup of spent fuel can be measured using burnup measurement equipment to compare and analyze the data. In fact, KHNP is measuring the burnup of spent fuel using the burnup measurement equipment (SICOM-NG-FA) developed by ENUSA in Spain. The burnup measurement equipment analyzes the axial burnup profile of spent fuel using gamma and neutron detectors. Burnup measurement is performed by moving the spent fuel up and down inside the measurement equipment and measuring the burnup of the fuel surface facing the gamma and neutron detectors. This paper aims to compare the results of measuring the burnup of spent fuel on two sides versus four sides using the burnup measurement equipment.

The Comprehensive Analyzer of Real Estimation for spent fuel POOL (CAREPOOL) has been developed for evaluating the thermal safety of a spent nuclear fuel pool (SFP) during the normal and accident conditions. The management of spent nuclear fuel function provides a management tool for spent nuclear fuel in the SFP. The fuel assemblies both in SFP and reactor side can be shown graphically in the screen. The loading sequence into transfer cask can be checked respectively in the CAREPOOL. A basic heat balance equation was used to estimate the SFP temperature using the heat load calculated in the previous step. The characteristics of typical SFPs and associated cooling systems at reactor sites in the Korea were applied. Accident simulation like station black out leading to loss of SFP cooling or inventory is possible. Emergency cooling water injection pipe installed subsequent to the events at Fukushima 2011 is also modeled in this system. The CAREPOOL provides four main functions- management of spent nuclear fuel, decay heat calculation by ORIGEN-S code, estimation of the time to boil/fuel uncovering by thermal-hydraulics calculations, fuel selection for periodic spent fuel transferring campaign. All of these are integrated into the GUI based CAREPOOL system. The CAREPOOL would be very beneficial to nuclear power plant operator and trainee who have responsibility for the SFP operation.

The dry storage of spent fuel has become an increasingly important issue in the field of nuclear energy. Square-gridded baskets have been widely used for the storage of spent fuel because of their superior heat transfer and structural integrity. In this paper, we review the fabrication process of square-gridded baskets for dry storage of spent fuel. The review includes the design considerations, material selection, manufacturing methods, and quality control measures. We also discuss the challenges and opportunities for further improvement in the fabrication of square-gridded baskets. The fabrication of square-gridded baskets is a critical process for the safe and reliable dry storage of spent fuel. The review of the fabrication process highlights the importance of design considerations, material selection, manufacturing methods, and quality control measures. Continued efforts to improve the fabrication process will help to ensure the safe and secure storage of spent fuel.

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

After spent fuel is stored in a dry storage container, it becomes difficult to obtain information on the fuel’s characteristics. As a result, it is necessary to identify the characteristics of spent nuclear fuel in advance and secure the information necessary to establish delivery acceptance requirements for interim storage and disposal in the future. Therefore, it is necessary to evaluate the characteristics of spent fuel before loading dry storage casks. In order to prepare for the dry storage of spent fuel, information on the basic characteristics of the fuel is required. As part of this information, it is also necessary to establish calculation criteria for spent fuel burnup. Spent fuel burnup can be classified into three categories. The first is burnup evaluated using design codes (design burnup), the second is burnup measured by furnace instruments during power plant operation (actual burnup), and the third is burnup measured through measurement equipment (measured burnup). This paper describes a comparative evaluation of design burnup, actual burnup, and measured burnup for specific fuels (40 bundles).

As of 2023, there has been significant progress worldwide in the management of nuclear fuel’s spent radioactive waste (HLW). Several countries have made important strides in advancing their plans for the construction of deep geologic repositories (DGRs) to safely dispose of their nuclear waste. Finland led the way, with its nuclear waste management organization, Posiva Oy, submitting an application for an operating license for a DGR for spent fuel generated by the nuclear power plants of its owners. The facility, ONKALO, will be located on the island of Olkiluoto and is expected to begin final disposal in the mid-2020s. Sweden also approved SKB’s application to build a DGR in Forsmark, and an encapsulation plant next to the Clab interim storage facility. In Switzerland, Nagra selected Nordic Lagern as the site for the Swiss DGR, and is preparing the general license applications for the required facilities. Meanwhile, Canada’s Nuclear Waste Management Organization (NWMO) narrowed down the possible locations for its DGR to two, and expects to name its preferred site by fall 2024. The UK established four Community Partnerships to participate in the siting process for a DGR, with Nuclear Waste Services (NWS) responsible for identifying a site. Andra, the French organization responsible for managing all French radioactive waste, is expected to submit an application by the end of the year for a DGR in France that will contain HLW resulting from reprocessing of spent fuel assemblies from French nuclear power plants, as well as intermediate-level waste. Overall, the progress made by these countries represents a tangible and sustainable step forward in the management of spent fuel and HLW, and brings us closer to the safe and effective long-term disposal of nuclear waste.

One of the most important factors in the delivery and acceptance requirements for dry storage of spent fuel is the burnup of spent fuel. Here, burnup has a unit of MWD/MTU and is used as a measure of how much nuclear fuel is depleted in a nuclear reactor. In addition, since it is one of the most basic characteristic information for the soundness evaluation of spent nuclear fuel, it is a required item not only by regulatory agencies but also by KORAD, the acquiring agency. The burnup of spent nuclear fuel is the burnup calculated through flux mapping using signals measured from in-reactor instruments during nuclear power plant operation (hereinafter: actual burnup) and the burnup calculated using the core design code (hereinafter: design burnup). In this paper, the design burnup of spent nuclear fuel discharged from OPR100 NPPs (Nuclear Power Plants) in Korea was recalculated to confirm the reliability of the actual burnup currently managed at the nuclear power plant. Basically, since spent nuclear fuel must maintain subcriticality under wet storage or dry storage, a burnup error of about 5% is considered as a conservative approach when evaluating the criticality safety of wet storage tanks and dry storage systems. Therefore, in this paper, we tried to verify whether the difference between actual burnup and design burnup for all spent nuclear fuel released from domestic OPR100 type light water reactor nuclear power plants is within 5%. As a result of the evaluation, the largest deviation between actual burnup and design burnup was about 1,457 MWD/MTU, and when converted into a percentage, it was about 3.3%. Therefore, it was confirmed that the actual burnup managed by OPR1000 NPPs in Korea has sufficient reliability. In the future, we plan to check the reliability of the performance burnup managed in WH NPPs, and some of them will be verified through measurement.

시설재배지 고추를 가해하는 꽃노랑총채벌레(Frankliniella occidentalis)를 대상으로 미끌애꽃노린재(Orius laevigatus)를 이용한 생물적 방 제가 검토되고 있다. 그러나 대상 해충의 빠른 집단 성장은 화학 살충제의 투입이 때에 따라 요구된다. 본 연구는 화학 살충제와 천적의 이상적 종합 방제를 구현하기 위한 목적으로 선택성이 높은 살충제 선발 및 이들 살충제 처리 이후 미끌애꽃노린재의 안전한 재투입 시기를 결정하기 위해 수행 되었다. 첫째로 꽃노랑총채벌레에 방제 효과가 높은 상용 살충제가 선발되었다. 총 17종류의 상용 살충제 가운데 5종류(pyriproxyfen+ spinetoram, abamectin, spinosad, acetamiprid, chlorpyrifos) 주성분을 갖는 상용 살충제가 꽃노랑총채벌레에 우수한 방제효과를 주는 약제로 선발되었다. 이들 5종류의 살충제에 대해서 미끌애꽃노린재의 감수성 반응은 꽃노랑총채벌레와 상이하였다. 특별히 아바멕틴과 스피네토람이 유 기인계 또는 네오니코티노이드에 비해 상대적으로 낮은 독성을 보였다. 이들 5종류의 살충제 처리 이후 잔류 독성을 미끌애꽃노린재를 이용하여 생물검정한 결과 유기인계 및 네오니코티노이드 약제는 비교적 오랜 기간 독성을 유지하지만, 아바멕틴과 스피네토람 약제의 경우 3일 이후에는 대상 천적에 피해를 주지 않는 것으로 나타났다. 이러한 잔류독성결과는 LC-MS/MS를 이용한 농약 잔류량 화학분석을 통해 뒷받침되었다. 이상 의 결과는 높은 밀도로 증가한 꽃노랑총채벌레에 대해서 이 해충에 살충성이 높은 아바멕틴 또는 스피네토람의 약제를 살포하고 이후 3일 지나 미 끌애꽃노린재의 투입을 통해 대상 해충의 평균 밀도를 경제적피해수준 이하로 유지할 수 있다는 종합방제 기술을 제시하고 있다.

In Korea, borated stainless steel (BSS) is used as spent fuel pool (SFP) storage rack to maintain nuclear criticality of spent fuels. As number of nuclear power plants and corresponding number of spent fuels increased, density in SFP storage rack also increased. In this regard, maintain subcriticality of spent nuclear fuels was raised as an issue and BSS was selected as structural material and neutron absorber for high density storage rack. Because it is difficult to replace storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and it is specified in the ASTM A887-89 standard depending on the boron concentration from 304B (0.20-0.29% B) to 304B7 (1.75-2.25% B). Due to low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr)2B are formed as secondary phase metallic borides could make Cr depletion near it which could decrease the corrosion resistance of material. In this paper, long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP condition. Because corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis was conducted with scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, hematite structure oxide film is formed and pitting corrosions occur on the surface of specimens. Most of pitting corrosions are found at the substrate surface because corrosion resistance of substrate, which has low Cr content, is relatively low. Also, oxidation reaction of B in the secondary phase has the lowest Gibbs free energy compared to other elements. Furthermore, oxidation of Cr has low Gibbs free energy which means that oxidation of B and Cr could be faster than other elements. Thus, the long-term corrosion might affect to boron content and the neutron absorption ability of the material.

In this study, surface roughness and interfacial defect characteristics were analyzed after forming a high-k oxide film on the surface of a prime wafer and a test wafer, to study the possibility of improving the quality of the test wafer. As a result of checking the roughness, the deviation in the test after raising the oxide film was 0.1 nm, which was twice as large as that of the Prime. As a result of current-voltage analysis, Prime after PMA was 1.07 × 10 A/cm2 and Test was 5.61 × 10 A/cm2, which was about 5 times lower than Prime. As a result of analyzing the defects inside the oxide film using the capacitancevoltage characteristic, before PMA Prime showed a higher electrical defect of 0.85 × 1012 cm2 in slow state density and 0.41 × 1013 cm2 in fixed oxide charge. However, after PMA, it was confirmed that Prime had a lower defect of 4.79 × 1011 cm2 in slow state density and 1.33 × 1012 cm2 in fixed oxide charge. The above results confirm the difference in surface roughness and defects between the Test and Prime wafer.