본고는 ‘작품의 통합성’이 왕부지의 시학의 중심을 관통하는 중요한 이론으로 보고, 이를 시 창작의 과정을 따라가며 어떻게 적용되는지 검토하였다. 곧 시 쓰기 전의 창작 관념에서 神의 운용과 現量을 중시하며 ‘하나의 뜻’[一意]이 주도하도록 하였으 며, 실제 시를 쓰는 제작 과정에서는 전편과 통합된 투탈어를 강조하였고, 작품 완성 후 감상할 때는 ‘하얀 바탕의 빛나는 비단’으로 비유되는, 자연스러운 통합성을 이룬 작품을 이상으로 여겼다. 여기에서 現量은 詩想의 단계에서만 작용하는 것이 아니라, 구성의 전과정에 관여하였다. 情과 景도 작품의 통합성이란 관점에서 상호 의존적인 개념으로 정립되었다. 渾成과 平美의 미학도 작품의 통합성이 전제되어야 가능한 사 실도 확인되었다. 창작론, 작품론, 형태론을 연결시켜 왕부지의 시학을 재구축하려 시도했다는 점에서 학계에 일정한 기여를 하리라 본다.

To safely dispose of highly radioactive spent resin and concentrate waste generated through nuclear power plant operations, it is essential to meet the physicochemical properties requirements of the packages and ensure the accuracy and reliability of radiological characteristics determination. Both spent resin and concentrate are packaged in high-integrity containers (HICs) after drying and are homogeneous waste products generated in the primary system and liquid radioactive waste treatment system. Meeting the physicochemical properties requirements does not appear to be difficult. However, to achieve reliable radiological characterization of high-integrity container packages, it is necessary to take a representative sample and perform accurate radiological analysis. Therefore, this paper discusses the methodology for evaluating the radionuclide inventory of high radioactive resin and concentrate packages, as well as the essential element technology and considerations. For relatively high radioactive resin and concentrate packages, the radionuclide inventory for each package should be evaluated with high reliability through direct radiological analysis of the representative samples collected for each package. This can contribute to the efficient operation of radioactive waste disposal facilities. Radionuclide-specific concentrations directly analyzed for each package will be managed in a database. As analytical data accumulates and direct measurements of high-integrity container package such as the radwaste drum assay system (RAS) become feasible, statistical techniques such as correlation analysis between easy-tomeasure (ETM) nuclides and difficult-to-measure (DTM) nuclides can lead to the development of efficient and reasonable indirect evaluation methods, such as scaling factor and the mean activity concentration method. As for the element technology, a remote representative sampling technique should be developed to safely and effectively take representative samples of highly radioactive materials, including granulated or hardened concentrate waste. Considerations should also be given to determining the sample quantity representing each package, as well as establishing radiation calibration and measurement methods appropriate to the radiation levels of the representative samples.

The operation time of a disposal repository is generally more than one hundred years except for the institutional control phase. The structural integrity of a repository can be regarded as one of the most important research issues from the perspective of a long-term performance assessment, which is closely related to the public acceptance with regard to the nuclear safety. The objective of this study is to suggest the methodology for quantitative evaluation of structural integrity in a nuclear waste repository based on the adaptive artificial intelligence (AI), fractal theory, and acoustic emission (AE) monitoring. Here, adaptive AI means that the advanced AI model trained additionally based on the expert’s decision, engineering & field scale tests, numerical studies etc. in addition to the lab. test. In the process of a methodology development, AE source location, wave attenuation, the maximum AE energy and crack type classification were subsequently studied from the various lab. tests and Mazars damage model. The developed methodology for structural integrity was also applied to engineering scale concrete block (1.3 m × 1.3 m × 1.3 m) by artificial crack generation using a plate jacking method (up to 30 MPa) in KURT (KAERI Underground Research Tunnel). The concrete recipe used in engineering scale test was same as that of Gyeongju low & intermediate level waste repository. From this study, the reliability for AE crack source location, crack type classification, and damage assessment increased and all the processes for the technology development were verified from the Korea Testing Laboratory (KTL) in 2022.

Due to the saturation of spent fuel pool of nuclear power plant in Korea, temporary storage for spent fuel will be installed, and spent fuel will be stored and managed in dry cask for a considerable period of time. Since spent nuclear fuel must withstand continuous decay heat, radiation and high internal pressure of the fuel rod in the cask, behavior of spent nuclear fuel is needed to be reviewed. Spent nuclear fuel used in Pressurized Water Reactor (PWR) in Korea is stored in a wet storage currently, but it is going to store a temporary dry-storage facility on Kori site. Therefore, it is very important and meaningful to evaluate the behavior of nuclear fuel with realistic modeling. Also, domestic PWR nuclear fuel has various burn-up. In the past, the burn-up of nuclear fuel in light water reactors was low, but in order to increase power generation efficiency, the concentration of uranium was increased and the number of new fuel was increased. Therefore, a large amount of nuclear fuel with burn-up of 45,000 MWD/MTU or higher, generally called high burn-up, is also stored in the spent fuel pool (SFP). Therefore, it is necessary to evaluate by dividing three different burn-up such as, low, medium, and high burn-up. Thus, this study will review the behavior of nuclear fuel at different burn-up during the temporary storage period with FALCON (EPRI), computational code and analyze the factors affecting the integrity of nuclear fuel, including when the temporary storage is extended its additional lifetime. And this evaluation will contribute developing the spent fuel management plan in Korea.

In KNF, fuel performance analysis modules were developed to predict the overall behavior of a fuel rod under normal operating conditions. Their main focus is to provide information on initial conditions prior to dry storage. Potential degradation mechanisms that may affect sheath integrity of spent CANDU fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed modules that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules, identify and extend the ranges of all modules to required operating ranges. The 300°C spent CANDU fuel sheath temperature metric for dry storage ensures spent CANDU fuel element integrity from the failure mechanisms of creep rupture, oxidation and stress corrosion cracking at a failure probability of 2×10-5 for a dry storage time of 100 years. The 300°C sheath temperature metric for dry storage has relatively a lower failure rate than the target criteria for dry storage of spent LWR fuel. Although different modes of failure were treated separately for simplicity, ignoring possible synergistic effects, these results are conservative because of the conservative assumptions that have been made for evaluating spent fuel element conditions, and because of the inherent conservatism of the applied models. Additional conservatism of the model comes from the fact that isothermal conditions do not prevail in actual storage conditions. Further R&D being considered includes acquisition of new functional models to implement overall fuel behavior evaluation and cover spent CANDU fuel in dry storage, and upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The developed modules provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for spent CANDU fuel.

A lot of CANDU Spent Fuels (CSFs) have been stored in spent nuclear fuel pools and dry storage facilities. In accordance with the enhanced nuclear regulations, the initial characteristics of CSF should be inspected to ensure the integrity of CSF and the reliable operation of storage system before loading it into a cask for long-term dry storage. For the inspections, an initial characteristics measurement equipment was designed, which is used for Pool-Side Examination (PSE) in the spent fuel pool of the pressurized heavy water reactor nuclear power plant. Measurements using the equipment consist of non-contact inspections and contact inspections. The non-contact inspections do not affect CSF integrity, whereas the integrity of CSF can be reduced during the contact inspections under abnormal operating conditions because the probe of equipment may apply specific loads to the CSF. Therefore, the structural integrity evaluations of equipment and CSF are performed using Finite Element (FE) analyses for four combinations based on two abnormal conditions and two probe positions. The used abnormal conditions are the pressing load condition and the scratching load condition, and two probe positions are the center and bottom of the fuel rod in the longitudinal direction, respectively. In this evaluation, the bottoms of the fuel rod or CSF are defined as the regions facing the bottom surface of equipment. The analysis of the pressing load condition is performed by pressing the probe of the equipment in radial direction of the CSF fuel rod. That of the scratching load condition is carried out by applying a specific radial load to the CSF fuel rod using the probe and then applying the load to the surface of the fuel rod while moving axially along the surface. All combinations are analyzed considering geometric, boundary and material non-linearity under the dynamic load, which is dependent on the equipment operating velocity. The stresses of CSF and equipment components were obtained from these analyses. The maximum stress of each component was generated at the combination on the scratching load condition for the bottom position among the four combinations. The obtained maximum stresses are lower than the yield stress for each component material. Also, the CSF is not overturned due to the support plate of the equipment in all analyses. Therefore, the structural integrity and safety of the equipment and the CSF are maintained under abnormal operating conditions during the inspection using the initial characteristic measurement equipment.

During the operation of a nuclear power plant (NPP), the generation of radioactive waste, including dry active waste (DAW), concentrates, spent resin, and filters, mandates the implementation of appropriate disposal methods to adhere to Korea’s waste acceptance criteria (WAC). In this context, this study investigates the potential use of polymer concrete (PC) as a high-integrity container (HIC) material for solidifying and packaging these waste materials. PC is a versatile composite material comprising binding polymers, aggregates, and additives, known for its exceptional strength and chemical stability. A comprehensive analysis of PC’s long-term integrity was conducted in this study. First, its compressive strength, which is crucial for ensuring the structural stability of HICs over extended periods, was evaluated. Subsequently, the resilience of PC was tested under various stress conditions, including biological, radiological, thermal, and chemical stressors. The findings of this study indicate that PC exhibits remarkable long-term properties, demonstrating exceptional stability even when subjected to diverse stressors. The results therefore underscore the potential viability of PC as a reliable material for constructing high-integrity containers, thus contributing to the safe and sustainable management of radioactive waste in NPPs.

Licensing for the application of the Polymer Concrete High Integrity Container (PC-HIC) to nuclear power plants has been completed or is in progress. Approval for the expanded application to all domestic nuclear power plants has been completed to utilize the 860 L PC-HICs for the 2nd stage surface repository, and the regulatory body is reviewing the license application to use the 510 L PCHICs for the 1st stage underground repository in the representative nuclear power plants. The 860 L PC-HICs, which have been licensed for all domestic nuclear power plants, will be used for safe storage management and disposal of low-dose dried concentrate waste and spent resin, and a total of 100 units is expected to be supplied to representative nuclear power plants that have been licensed first. The 510 L PC-HICs are planned to be used for underground disposal of high-dose spent resin and dried concentrate waste. Prior to the application of PC-HICs to nuclear power plants and disposal to the repository, it is necessary to establish realistic and reasonable requirements through close consultations between waste generator and disposal operators to ensure the suitability for disposal of PC-HIC packages and to carry out disposal delivery and acceptance work. Since the Polymer Concrete High Integrity Container (PC-HIC) has long-term integrity of more than 300 years and the barrier does not temporarily collapse, spent resin and dried concentrate waste, which are radioactive wastes to be solidified, can be disposed of much more safely in PC-HIC packages than solidified types. Acceptance criteria for the PC-HIC packages should be prepared fully reflecting the advantages of PC-HIC, and quality assurance methods for physical/chemical/radiological characterization results based on the Waste Certification Program (WCP) should be supported. In addition, infrastructure should be secured for safe transportation, handling, and storage of the PC-HIC packages. In this paper, we have tried to find a reasonable acceptance criteria, quality assurance method, and infrastructure level according to the dose and disposal conditions of PC-HIC packages.

Plasma torch melting has been considered as a promising treatment technology for radioactive waste generated by nuclear power plants. The IAEA reported in 2006, the plasma melting technology could be treated regardless of the type of radioactive wastes such as combustible, non-combustible and liquid. Also, the technology has the advantage of being an eco-friendly technology. It emits less harmful gases such as NOx, SOx, HCl and CO because it does not use fossil fuels. In KHNP CRI, the plasma torch melting system was developed as the new radioactive waste treatment technology. In this study, to evaluate the long-term integrity of the new facility, a demonstration test with concrete as a simulant was carried out for about 3 days. For the 3 days, the evaluation was conducted in view of abnormal shutdown, soundness of waste feeding device, electrode consumption, and so on.

CANDU Spent Fuel (CSF) dry storage system, SILO, has been operated from 1992 at Wolsung under 50 year operating license. As of 2023, this system has been operated for over 30 years and its licensed remaining operation time is less than 20 years. When it faces the final stage of operation, it has only two options; moving to a centralized away-from-reactor storage or extending its license atreactor. These two options have an inevitable common duty of confirming the CSF integrity by a “demonstration test”. Since the degradation of CSF and structural materials in the SILO are critically dependent on temperature, two important goals of the ‘DEMO test’ were set as follows. 1. Design of ‘DEMO SILO’: Development of internal monitoring technology by transforming SILO design. 2. Accurate measurement and evaluation of the three-dimensional temperature distribution in the ‘DEMO SILO’ Based on operating real commercial SILO dimension, a conceptual “DEMO SILO” design has been developed from 2022. Because, unlike with commercial Silo, ‘Demo Silo’ must be disassembled and assembled, and have penetration holes. Safety evaluation technologies like structural, thermal and radiation protection analysis also have been developed with design work. ‘Demo SILO’ should evaluate an accurate 3D temperature distribution with minimal number of thermocouples and penetration holes to avoid disruption of internal flow and temperature distribution. For this reason, a ‘Best Estimate Thermal-Hydraulics evaluation system for SILO’ is under development and it will be essential for ensuring temperature prediction accuracy. Construction of a full-scale test apparatus to validate this technology will begin in 2024. In order to supply power to many heaters and monitor temperature gradient inside of this apparatus, it has modular design concept by dividing its whole body to axial 9 sub-bodies which looks like a donut containing a basket at center position.

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

Spent fuel from the Wolsong CANDU reactor has been stored in above-ground dry storage canisters. Wolsong concrete dry storage canisters (silos) are around 6 m high, 3 m in outside diameter, and have shielding comprised of around 1 m of concrete and 10 mm of steel liner. The storage configuration is such that a number of fuel bundles are placed inside a cylindrical steel container known as a Fuel Basket. The canisters hold up to 9 baskets each that are 304 L stainless steel, around 42” in diameter, 22” in height, and hold 60 fuel bundles each. The operating license for the dry storage canisters needs to be extended. It is desired to perform in-situ inspections of the fuel baskets to very their condition is suitable for retrieval (if necessary) and that the temperature within the fuel baskets is as predicted in the canister’s design basis. KHNP-CNL (Canadian Nuclear Lab.) has set-up the design requirements to perform the in-situ inspections in the dry storage canisters. This Design Requirements applies to the design of the dry storage canister inspection system.

Integrity evaluation scheme for Spent Fuel (SF) dry storage has been developed under transportation failure modes. This method especially considered the degradation characteristics of Spent Fuel (SF) during dry storage such as radial and circumferential hydride content, hydride volume fraction, oxide thickness, etc. Hydride and zircaloy cladding are considered as material composite system, using correlation models related to material properties. Critical Strain Energy Density (CSED) is compared with Strain Energy Density (SED), to evaluate cladding integrity. CSED serves as material characteristics, while SED can be considered as boundary condition. To calculate the CSED of cladding in the lateral failure mode, circumferential hydride concentration is used. SED is calculated considering both the bending moment and axial load. On the other hand, in the longitudinal failure case, fuel rod temperature, internal pressure, hoop stress, radial hydride concentration is used to calculate CSED. And pinch force (contact) was considered to evaluate SED. Model validations were conducted by comparing hot cell SF test and existing validated evaluation results. To separately handle normal transportation conditions from hypothetical accident conditions, SED according to stress-strain analysis results was separated into elastic and plastic regions. As a result of applying this scheme for 14×14 SF, failure probability of normal condition was zero, which is the similar result with DOE and same with EPRI. Regarding accident condition, lateral case showed similar result, but longitudinal case showed different but reasonable result, which was due to the different analysis conditions. The proposed methodology which was indigenously developed through this study is named as K-method.

Research ethics is a desirable way of conducting responsible research, which means maintaining the integrity and complying with ethical norms in a particular field of research. Research integrity means keeping the values and standards required by researchers in the research process. Despite educational institutions' efforts to improve research ethics, the judgment of research cheating is on the rise. In addition, there were unclear regulations related to self-plagiarism, duplicate publication, and the scope of research ethics. If you are a researcher of human research, planning and conducting ethical research is the first step for research integrity. Researchers need to understand protecting research subjects conducting research, and educational institutions should review and improve the current curriculum to provide research ethics education more effectively. Before strengthening legal sanctions on research cheating centered on results, the government should seek solutions through fundamental cause analysis to prevent researchers from committing research cheating. Given the current status of research ethics educational institutions and the progress of policies so far, we hope that if everyone tries at the individual, educational institution, and government levels, we will move in a more positive direction.

This study was conducted to evaluate the effects of ImmunoSEB as an immune-booster additive on the performance of broiler chicks. A total of 1150 broiler chickens were randomly allotted to three treatments on the basis of body weight (10 pens per treatment with 40 broilers in each pen): the control (CON), CON + ImmunoSEB 0.025% in feed (SEB25), and CON + ImmunoSEB 0.050% in feed (SEB50). The experiment was conducted for d 42 in three phases (phase 1, d 0–14; phase 2, d 15–28; and phase 3, d 29–42). There were significant differences in the average daily gain (ADG) and feed intake. The ADG at d 14 in the SEB50 treatment was greater than that in the CON treatment. The overall ADG in the SEB50 treatment was greater than that in the CON treatment. During d 0–14, the feed intake of chickens in the SEB50 treatment increased compared to that in the CON treatment. The crude protein and lysine digestibility improved in the SEB25 and SEB50 treatments compared to those in the CON treatment at d 28. Superoxide dismutase concentration significantly increased in the SEB50 treatment compared to that in the CON treatment. The interleukin-1β and tumor necrosis factor-α concentrations were higher in the cecum of chickens in the CON treatment than in the SEB25 and SEB50 treatments. A lower population of E. coli was detected in the ileum and cecum of broilers fed the SEB50 diet compared to those of broilers fed the CON diet. The overall result showed the beneficial effects of using ImmunoSEB at a dose of 0.050% in broiler chickens.

본 연구는 반부패·청렴의 개념화를 분류하고 측정하기에 앞서 관련 정책이 생성‧발생하고 작동하는 원리 자체로, 반부패·청렴의 개념화를 어떻게 인지하고 이를 활용하고 있는지를 탐색한다. 부연하면, 4가지의 반부패와 청렴의 언어철학적 분석을 통해 맥락을 파악하고(①동의어, 이의어(②구분형, ③범주형, ④인과형), 현 반부패‧청렴정책에서 파악한 맥락이 어떻게 전제되어 있는가를 살펴본다. 이를 통해, 반부패·청렴의 정책문제에서 의미문제와 제도수립의 문제로, 반부패·청렴의 개념화로 인한 상황으로부터의 단절, 개념화와 재량권의 발생, 기술의 서술적 한 계와 재량권, 의사소통의 가능성 문제, 이중 해석 및 번역의 문제 등으 로 인해 거래비용이 발생하는 점 등을 정책에 반영해 합리적이고 효율 적으로 성공적인 정책이 되도록 해야 함을 제안한다.

The effect of permanganate oxidation was investigated as water treatment strategy with a focus on comparing the reaction characteristics of NaOCl and sodium permanganate (NaMnO4) in algae (Monoraphidium sp., Micractinium inermum, Microcystis aeruginosa)-contained water. Flow cytometry explained that chlorine exposure easily damaged algae cells. Damaged algae cells release intracellular organic matter, which increases the concentration of organic matter in the water, which is higher than by NaMnO4. The oxidation reaction resulted in the release of toxin (microcystin-LR, MC-LR) in water, and the reaction of algal organic matter with NaOCl resulted in trihalomethanes (THMs) concentration increase. The oxidation results by NaMnO4 significantly improved the concentration reduction of THMs and MC-LR. Therefore, this study suggests that NaMnO4 is effective as a pre-oxidant for reducing algae damage and byproducts in water treatment process.

Present study investigated the waste form integrity of melted products generated from PAM-MSO system, which is proposed and developed to compensate the drawbacks of each system. The disposal suitability of the melting solidification products generated from the plasma arc melting treatment of pulverized cement debris spiked by Pb, Cd and Cs, as indicators of typical hazardous metals and radionuclides existed in the low-level mixed waste in the KHNPPs. The final waste form obtained by the test was evaluated for suitability for disposal. The compressive strength was 261.10 MPa, showing much higher values when compared to other waste form products. The compressive strength of both the sample after irradiation with 107 Gy radiation and that after long-term submersion test (90 days) satisfied the disposal criteria. As a result of the leaching test conducted according to the ANS 16.1 test method, it was confirmed that the leaching index satisfies the disposal criteria.

Radioactive spent resin and concentrate waste powder generated from the primary system of nuclear power plants (NPPs) should be treated and disposed of in the form of solidified products or high integrity container (HIC) packages. We are preparing for the application of polymer concrete high integrity containers (PC-HICs) that has been approved for disposal and field application after going through the disposal suitability review of the repository operator and the license review process of the regulatory body. A reliable assessment of nuclide inventory in waste drum is required for the disposal of the radioactive waste drums, and the representative samples should be collected for both the indirect (non-destructive assessment based on the scaling factor, average radioactivity concentration, etc.) and direct (destructive analysis) evaluation of the difficult-to-measure (DTM) nuclides. It is important to secure the representativeness of samples for reliable and accurate evaluation of radionuclide inventory and approval of methodologies for highly radioactive waste such as spent resin and concentrate waste poser, and in order to secure the radiation safety of the sampling workers and representativeness of the samples, a remote sampling method is required with excellent convenience and safety and sufficient representativeness of the sample. The simple sampling device used in the past to collect samples for the scaling factor does not have a remote control function, so high-radiation samples must be collected within a very short time and it is difficult to obtain sufficiently representative samples due to structural characteristics that cannot collect the entire sample in the axial direction of the package. Therefore we developed concept designs for a remote sampling device that can satisfy both sample representativeness, operator convenience and safety.

High Integrity Container (HIC) made of polymer concrete was developed for the efficient treatment and safe disposal of radioactive spent resin and concentrate waste in consideration of the disposal requirements of domestic disposal sites. Permission for application of Polymer Concrete High Integrity Container (PC-HIC) to the domestic nuclear power plants (NPPs) has been completed or is under examination by the regulatory agency. In the case of 860 L PC-HIC for very-low-level-waste (VLLW) or low low-level-waste (LLW), the application of four representative NPPs has been approved, and the license for extended application to the rest NPPs is also almost completed. A licensing review is also underway to apply 510 L PC-HIC for intermediate and low-level-waste (ILLW) to representative nuclear power plants. In order to handle and efficiently store and manage PC-HICs and high-dose PCHIC packages, a gripper device that can be remotely operated and has excellent safety is essential, and the introduction of NPPs is urgent. The conventional gripper device developed by the PC-HIC manufacturer for lifting test to evaluate the structural integrity of PC-HIC requires a rather wide storage interval due to its design features, and does not have a passive safety design to handle heavy materials safely. In addition, work convenience needs to be reinforced for safety management of high radiation work. Therefore, we developed a conceptual design for a gripper device with a new concept to minimize the work space by reflecting on-site opinions on the handling and storage management conditions of radioactive waste in NPPs, and to enhances work safety with the passive safety design by the weight of the package and the function of checking the normal seating of the device and the normal operation of the grip by the detector/indicator, and to greatly improves the work efficiency and convenience with the wireless power supply function by rechargeable battery and the remote control function by camera and wireless monitoring & control system. Through design review by experts in mechanical system, power supply and instrumentation & control fields and further investigations on the usage conditions of PC-HICs, it is planned to facilitate preparations for the application of PC-HIC to domestic NPPs by securing the technical basis for a gripper device that can be used safely and efficiently and seeking ways to introduce it in a timely manner.