• Nuclear Engineering and Technology
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• 제43권2호
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.

• Journal of Radiation Protection and Research
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• 제37권4호
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• pp.197-201
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• 2012

경수로 사용후핵연료 중간저장시설의 부지면적을 산출하기 위하여 콘크리트 저장시설 개념모델의 연간선량을 계산하였다. 초기농축도 4.5 wt%, 연소도 45,000 MWd/MTU, 냉각기간 10년인 사용후핵연료를 대상으로 ORIGEN-ARP를 사용하여 선원항을 생산하였으며, MCNP 코드를 사용하여 저장시설에 대한 모델링 및 방사선차폐계산을 수행하였다. 연간선량은 저장시설의 용량별로 계산하였으며, 중앙집중식 저장시설의 경우, 반경 700 m 이상에서 10CFR72에서 권고하는 통제구역 경계에서의 연간선량 기준 0.25 mSv를 만족하였다.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.969-978
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• 2013

A full-sized model for the horizontally oriented metal cask containing 21 spent fuel assemblies has been considered to evaluate the internal natural convection behavior within a dry shield canister (DSC) filled with helium as a working fluid. A variety of two-dimensional CFD numerical investigations using a turbulent model have been performed to evaluate the heat transfer characteristics and the velocity distribution of natural convection inside the canister. The present numerical solutions for a range of Rayleigh number values ($3{\times}10^6{\sim}3{\times}10^7$) and a working fluid of air are further validated by comparing with the experimental data from previous work, and they agreed well with the experimental results. The predicted temperature field has indicated that the peak temperature is located in the second basket from the top along the vertical center line by effects of the natural convection. As the Rayleigh number increases, the convective heat transfer is dominant and the heat transfer due to the local circulation becomes stronger. The heat transfer characteristics show that the Nusselt numbers corresponding to $1.5{\times}10^6$ < Ra < $1.0{\times}10^7$ are proportional to 0.5 power of the Rayleigh number, while the Nusselt numbers for $1.0{\times}10^7$ < Ra < $8.0{\times}10^7$ are proportional to 0.27 power of the Rayleigh number. These results agreed well with the trends of the experimental data for Ra > $1.0{\times}10^7$.

• 방사성폐기물학회지
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• 제17권4호
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• pp.405-418
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• 2019

국내 원자력발전소에서 발생하는 사용후핵연료의 제원 및 방출시점 등 특성과 현재의 고준위 방사성폐기물 기본계획에 근거한 처분시나리오를 도출하여 기존 심층 처분시스템을 바탕으로 처분효율과 경제성을 향상시킨 개선된 처분시스템을 제안하였다. 이를 위하여 국내 원자력발전소에서 발생하는 사용후핵연료의 길이에 따라 2종류의 처분용기 개념을 도출하고, 사용후핵연료 발생 년도와 현재의 기본계획에 근거한 처분 시나리오 설정에 따른 처분시점에서의 냉각기간을 고려하여 처분용기내 수용 가능한 붕괴열 량을 결정하였다. 그리고 2종류의 처분용기에 대한 처분시스템과 결정된 붕괴열을 바탕으로 열적 안정성 분석을 통하여 제안된 처분시스템의 설계요건에 대한 적합성 여부를 확인하고, 처분효율을 평가하였다. 개선된 처분시스템은 기존 처분시스템에 비하여 처분면적은 약 20% 감소되고 처분밀도는 약 20% 향상됨을 확인하였고, 처분용기와 완충재 재료도 상당량 절감됨을 확인하였다. 본 연구의 결과는 향후 사용후핵연료 관리정책 수립 및 실제 사업을 위한 처분시스템 설계를 위한 자료로 활용될 수 있다.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2002년도 춘계학술대회 논문집
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• pp.840-843
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• 2002

The dry pulverizing/Mixing device is used to deal with the spent fuels for the safe disposal. The separated pellets from hulls by a slitting device are put and oxidized from UO$_2$ solid pellet to U$_3$O$\_$8/ powder in the device. The device have been developed based on a voloxidation method which is one of several dry de-cladding methods. We have benchmarked dry de-cladding methods, analyzed applicability to the advanced spent fuel management process, integrated and compared several configuration, and finally derived detailed specifications proper to requirements for the device. Also, thermal characteristics of the device such as thermal stress and strain have been analyzed by the commercial software, 1-DEAS, and the reliability of the results have been verified by the KOLAS(Korea Laboratory Accreditation Scheme). The UO$_2$ solid pellets are put in the device which has a capacity of 20 kgHM per a batch, heated up about 600$^{\circ}C$ in the air environment. Then, the UO$_2$ solid pellets are oxidized into the U$_3$O$\_$8/ powder, and the powder is collected in a special vessel. The device has been designed and developed as fellows: the multi-staged fine hole meshes are used to reduce the size of the powder gradually, heat and air(oxygen) are supplied continuously to reduce the reaction time, and slight vibration effect are applied to collect powder cling to the device.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1991년도 한국자동제어학술회의논문집(국내학술편); KOEX, Seoul; 22-24 Oct. 1991
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• pp.832-837
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• 1991

After many years of proliferation, the nuclear industry is indebted for a formidable consequence, the safe management of spent fuel. Naturally, the high radioactivity involved with such process motivates the development of effective telerobotic systems. Nevertheless, the existing master-slave type of tele manipulators are limited in effectiveness by the human operator's limited sensory and manipulation capabilities. This paper presents the result of a research effort to resolve such problems by assigning the slave manipulator a certain degree of intelligence; sensing and actuation. In the presented system, a perception-action loop is achieved using ultrasonic range sensor and laser distance sensor interfaced with the PUMA 760 industrial robot system, and applied to automating impact wrenching task for unbolting the lid of nuclear spent fuel cask. The perception-action loop performs determination of the cask location, collision avoidance and centering of the impact wrench onto the bolt head. To aid the insertion task and to provide versatility a mounting module consisting of an RCC device and an automatic tool changer is designed and implemented. The performance of the developed system is tested on the model cask and the result is given.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 1997년도 추계학술대회 논문집
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• pp.743-746
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• 1997

The Master-Slave manipulator is the generally used remote handling equipment in the hot cell, in which the high level radioactive materials such as spent fuels are handled. To analyze the motion and to implement the training system by virtual reality technology, the simulator for M-S manipulator using the computer graphics is developed. The parts are modelled in 3-D graphics, assembled, and kinematics are assigned. The inverse kinematics of the manipulator is defined, and the slave of manipulator is coupled with master by the manipulator's specification. Also, the virtual workcell is implemented in the graphical environment which is the same as the real environment. This graphic simulator of manipulator can be effectively used in designing of the maintenance processes for the hot cell equipment and enhance the reliability of the spent fuel management.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.142-147
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• 2021

Hydrogen peroxide is a radiolysis product of water formed under gamma-irradiation; therefore, its reliable detection is crucial in the nuclear industry for spent fuel management and coolant chemistry. This study proposes an electrochemical sensor for hydrogen peroxide detection. Cysteamine (CYST), gold nanoparticles (GNPs), and horseradish peroxidase (HRP) were used in the modification of a gold electrode for fabricating Au/CYST/GNP/HRP sensor. Each modification step of the electrode was investigated through electrochemical and physical methods. The sensor exhibited strong sensitivity and stability for the detection and measurement of hydrogen peroxide with a linear range of 1-9 mM. In addition, the Michaelis-Menten kinetic equation was applied to predict the reaction curve, and a quantitative method to define the dynamic range is suggested. The sensor is highly sensitive to H₂O₂ and can be applied as an electrochemical H₂O₂-sensor in the nuclear industry.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.411-420
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• 2003

원자력발전소 운전과 함께 국내 원전에서 계속 발생, 저장하고 있는 사용후핵연료를 안전하고도 효율적으로 관리하기 위하여 차세대관리 종합공정을 개발하고 있으며, 이를 실증하기 위한 시설을 설계 중에 있다. 이 실증시설은 조사재시험시설에 마련된 예비 핫셀을 차세대관리 종합공정의 특성을 고려하여 개조하여 사용할 예정이다. 이 연구에서는 실증시설에 대한 기존 시설 및 부대시설의 개조방안 등 기본 건축구조 설계에 대한 기준과 설계내용을 제시하였으며, 건축구조물의 안전성을 입증하기 위한 해석을 수행하고 그 결과를 제시하였다. 본 연구결과는 차세대관리 종합공정 실증시선의 상세설계를 위한 자료로 사용될 것이며, 시설의 인허가를 위한 자료로 활용될 것이다.