The PoN (Proof of Nonce) distributed consensus algorithm basically uses a non-competitive consensus method that can guarantee an equal opportunity for all nodes to participate in the block generation process, and this method was expected to resolve the first trilemma of the blockchain, called the decentralization problem. However, the decentralization performance of the PoN distributed consensus algorithm can be greatly affected by the network transaction transmission delay characteristics of the nodes composing the block chain system. In particular, in the consensus process, differences in network node performance may significantly affect the composition of the congress and committee on a first-come, first-served basis. Therefore, in this paper, we presented a problem by analyzing the decentralization performance of the PoN distributed consensus algorithm, and suggested a fairness control algorithm using a learning-based probabilistic acceptance rule to improve it. In addition, we verified the superiority of the proposed algorithm by conducting a numerical experiment, while considering the block chain systems composed of various heterogeneous characteristic systems with different network transmission delay.

In recent years, importance of blockchain systems has been grown after success of bitcoin. Distributed consensus algorithm is used to achieve an agreement, which means the same information is recorded in all nodes participating in blockchain network. Various algorithms were suggested to resolve blockchain trilemma, which refers conflict between decentralization, scalability, security. An algorithm based on Byzantine Agreement among Decentralized Agents (BADA) were designed for the same manner, and it used limited committee that enables an efficient consensus among considerable number of nodes. In addition, election of committee based on Proof-of-Nonce guarantees decentralization and security. In spite of such prominence, application of BADA in actual blockchain system requires further researches about performance and essential features affecting on the performance. However, performance assessment committed in real systems takes a long time and costs a great deal of budget. Based on this motivation, we designed and implemented a simulator for measuring performance of BADA. Specifically, we defined a simulation framework including three components named simulator Command Line Interface, transaction generator, BADA nodes. Furthermore, we carried out response surface analysis for revealing latent relationship between performance measure and design parameters. By using obtained response surface models, we could find an optimal configuration of design parameters for achieving a given desirable performance level.

Recently, blockchain technology has been recognized as one of the most important issues for the 4th Industrial Revolution which can be represented by Artificial Intelligence and Internet of Things. Cryptocurrency, named Bitcoin, was the first successful implementation of blockchain, and it triggered the emergence of various cryptocurrencies. In addition, blockchain technology has been applied to various applications such as finance, healthcare, manufacturing, logistics as well as public services. Distributed consensus algorithm is an essential component in blockchain, and it enables all nodes belonging to blockchain network to make an agreement, which means all nodes have the same information. For example, Bitcoin uses a consensus algorithm called Proof-of-Work (PoW) that gives possession of block generation based on the computational volume committed by nodes. However, energy consumption for block generation in PoW has drastically increased due to the growth of computational performance to prove the possession of block. Although many other distributed consensus algorithms including Proof-of-Stake are suggested, they have their own advantages and limitations, and new research works should be proposed to overcome these limitations. For doing this, above all things, we need to establish an evaluation method existing distributed consensus algorithms. Based on this motivation, in this work, we suggest and analyze assessment items by classifying them as efficiency and safety perspectives for investigating existing distributed consensus algorithms. Furthermore, we suggest new assessment criteria and their implementation methods, which can be used for a baseline for improving performance of existing distributed consensus algorithms and designing new consensus algorithm in future.

본 논문에서는 이단계 칼만필터를 활용한 구조물의 3 자유도 동적변위 계측 시스템을 소개한다. 개발 시스템은 센서 모듈, 베이스 모듈, 컴퓨테이션 모듈로 구성되어 있다. 센서 모듈은 100Hz 샘플주파수의 고정밀 가속도를 계측하는 포스피드백 가 속도계와 10Hz의 샘플주파수의 저정밀도의 속도, 변위를 계측하는 저가의 RTK-GNSS로 구성되어 있다. 계측된 데이터는 LAN 케이블을 통하여 컴퓨테이션 모듈로 전송되고, 컴퓨테이션 모듈에서 이단계 칼만필터를 활용하여 100Hz 샘플주파수의 고정밀 변위를 실시간으로 산정한다. 개발 시스템의 변위 계측 정밀도를 검증하기 위해 미국, 캘리포니아에 위치한 San Francisco-Oaklmand Bay bridge 에서 현장 실험을 수행하였으며, 실험 결과 1.68mm RMS 오차를 보임을 확인하였다.

The paper presents a new short-term dynamic displacement estimation method based on an acceleration and a geophone sensor. The proposed method combines acceleration and velocity measurements through a real time data fusion algorithm based on Kalman filter. The proposed method can estimate the displacement of a structure without displacement sensors, which is typically difficult to be applied to earthquake or fire sites due to their requirement of a fixed rigid support. The proposed method double-integrates the acceleration measurement recursively, and corrects an accumulated integration error based on the velocity measurement, The performance of the proposed method was verified by a lab-scale test, in which displacement estimated by the proposed method are compared to a reference displacement measured by laser doppler vibrometer (LDV).

Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.

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.

Regulatory agencies require burn-up verification to ensure that dry storage casks using burn-up credit are not loaded with fuel with a reactivity greater than the allowable standard. Accordingly, in preparation for dry storage of SF, the reliability of the burnup was verified and action plans for fuel with confirmed errors were reviewed. Reliability verification was performed by comparing the actual burnup calculated with combustion calculation code (TOTE, ISOTIN) used in NPP and the design burnup calculated with the nuclear design code (ANC). As a result of comparing the differences between actual burnup and design burnup for 7,414 assemblies of SF generated from CE-type NPPs, the average deviation was confirmed to be 0.79% and 220 MWD/MTU. In the CE-type NPPs, no fuel showing large deviations was identified, and it was confirmed that reliability was secured. As a result of comparing the differences in 11,082 assemblies of SF generated from WH-type NPPs, the differences were not large, averaging 1.16% or 422 MWD/MTU. However, fuels showing significant differences were identified, and cause analysis was performed for those fuels. The cause analysis used a method of comparing the burnup of symmetrically loaded fuels in the reactor. For fuels that were not symmetrically loaded, a method was used to compare them with fuels with similar combustion histories. As a result of the review, it was confirmed that the fuel was under- or over-burned compared to symmetrically loaded fuel. For fuels for which clear errors have been identified, we are considering replacing them with the design burnup, and for fuels whose causes cannot be confirmed, we are considering ways to maintain the actual burnup.

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.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

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.