Development of the Vacuum Drying Process for the PWR Spent Nuclear Fuel Dry Storage

Chang-Yeal Baeg,Chun-Hyung Cho 한국방사성폐기물학회 2016 방사성폐기물학회지 Vol.14 No.4

본 논문은 국내 원전의 습식저장조에 저장 중인 경수로형 사용후핵연료를 금속겸용용기를 이용해 건식으로 운영하기 위한운영공정을 개발하는 것이다. 국내 경수로형 원전의 사용후핵연료는 1990년대 초부터 습식으로 소내에서 운반을 한 경험은 많으나 건식으로 운전한 경험은 전혀 없는 실정이다. 이에 따라 금속겸용용기를 운영할 수 있는 세부 운영공정을 개발하였으며 주요 운영공정에서 금속겸용용기의 주요 구성품 및 사용후핵연료의 안전성이 유지됨을 확인하였다. 단기운영공정은 총 21시간 내에 이루어지도록 절차를 수립하였고 단계별로 허용운전 시간(15시간 습식공정, 3시간 배수공정, 그리고 3시간 진공공정)도 제시하였다. This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process).

Establishment of the Design and Testing Methods of Lid Welds as the Confinement Boundary for Spent Nuclear Fuel Storage Canister

Chang-Yeal Baeg,Chun-Hyung Cho 한국방사성폐기물학회 2017 한국방사성폐기물학회 학술논문요약집 Vol.2017 No.10

Development Status for Commercialization of Spent Nuclear Fuel Transportation and Dry Storage System Technology

Chang-Yeal Baeg,Chun-Hyung Cho 한국방사성폐기물학회 2018 방사성폐기물학회지 Vol.16 No.2

During the seven years from 2009 to 2016, PWR SNF (spent nuclear fuel) transportation and storage systems suitable for domestic conditions were developed by the government to cope with the saturation of wet storage capacity in NPPs. One of the developed systems is a multipurpose metal cask applicable for transportation/storage; the other is a concrete cask dedicated to storage. Efficient cask technologies were secured utilizing the characteristics and experience of relevant industrial, academic and research institutes. Technological independence was also achieved through several patent registrations of research outcomes. To prepare for a rapid increase of demand in the near future, technology transfer of secured patents and technologies to the domestic industry was carried out twice in the years of 2016 and 2017. This

Design Concept of Package Stowage and Retention Systems During the Road and Sea Transport

Chang-Yeal Baeg,Sang-Jin Lee 한국방사성폐기물학회 2018 한국방사성폐기물학회 학술논문요약집 Vol.2018 No.10

On-Site Transport and Storage of Spent Nuclear Fuel at Kori NPP by KN-12 Transport Cask

Sung-Hwan Chung,Chang-yeal Baeg,Byung-Il Choi,Ke-Hyung Yang,Dae-Ki Lee 한국방사성폐기물학회 2006 방사성폐기물학회지 Vol.4 No.1

고리 원전 사용후핵연료 저장조의 저장용량을 확보하기 위하여 2002년부터 사용후핵연료 운반용기를 이용하여 400다발 이상의 PWR 사용후핵연료 집합체를 원전부지 내에 수송, 저장하였다. 이를 위하여 KN-12 운반용기, 관련장비 및 수송차량으로 구성되는 수송시스템을 구성하였다. KN-12 운반용기는 국내 원자력법 및 IAEA의 수송규정에 따라 설계, 제작되고, 정부로부터 인허가를 획득하였으며, 취급장비 역시 관련규정에 따라 구비하였다. 수송 저장작업은 2 대의 운반용기를 동시에 투입하여 수행하였으며, 모든 작업공정에 대하여 엄격한 품질관리 및 방사선 안전관리를 수행하여 수송 안전성을 확보하고 신뢰도를 제고하였다. Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA's transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.

Structural Safety Analysis of Lifting Device for Spent Fuel Dual-purpose Metal Cask

Tae-Chul Moon,Chang-Yeal Baeg,Si-Tae Yun,Byung-il Choi,In-su Jung 한국방사성폐기물학회 2014 방사성폐기물학회지 Vol.12 No.4

인양장비는 원자력발전소에서 발생하는 사용후핵연료를 운반하는 운반용기를 인양하기 위해 사용된다. 본 연구는 원자력 안전위원회고시 제2013-27호와 미국 10CFR Part 71 §71.45에서 규정하는 기술수준에 따라 이론적인 방법과 유한요소방법 으로 인양장비의 구조적안전성을 평가하였다. 이론적으로 평가한 결과 모든 구성 요소에서의 응력이 응력제한치 내에 있어 운영상 발생하는 구조적 안전성을 확보하고 있는 것으로 판단하였다. 또한 유한요소해석을 통한 평가결과, 항복과 극한조건 모두에서 설계기준을 만족하는 것으로 평가되었다. 모든 구성요소에서 충분한 안전여유도(항복조건에서 3 이상의 안전율, 극한조건에서 5 이상의 안전율)를 갖는 것으로 나타나 구조적으로 안전하다고 판단하였다. A lifting device is used to deal with transport cask for the transportation of spent fuels from nuclear power plants. This study performed theoretical analysis and numerical simulation to evaluate the structural integrity of the lifting device based on Nuclear Safety and Security Commission(NSSC) Notice No.2013-27 and US 10CFR Part 71 §71.45. The results of theoretical analysis showed that the maximum stresses of all components were below the allowable values. This result confirmed that the lifting device was structurally safe during operation.The results of finite element analysis also showed that it was evaluated to satisfy the design criteria bothyielding and ultimate condition. All components have been shown to ensure the structural safety due to sufficient safety margins. In other words, the safety factor was 3 or more for the yielding condition and was 5 or more for the ultimate condition.

Structural Evaluation under Off-normal Condition for Concrete Storage Cask of Spent Fuels

Tae-Chul Moon,Chang-Yeal Baeg,Chun-Hyung Cho 한국방사성폐기물학회 2017 한국방사성폐기물학회 학술논문요약집 Vol.2017 No.05

SHIELDING ANALYSIS OF DUAL PURPOSE CASKS FOR SPENT NUCLEAR FUEL UNDER NORMAL STORAGE CONDITIONS

Ko, Jae-Hun,Park, Jea-Ho,Jung, In-Soo,Lee, Gang-Uk,Baeg, Chang-Yeal,Kim, Tae-Man Korean Nuclear Society 2014 Nuclear Engineering and Technology Vol.46 No.4

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

NATURAL CONVECTION HEAT TRANSFER CHARACTERISTICS IN A CANISTER WITH HORIZONTAL INSTALLATION OF DUAL PURPOSE CASK FOR SPENT NUCLEAR FUEL

Lee, Dong-Gyu,Park, Jea-Ho,Lee, Yong-Hoon,Baeg, Chang-Yeal,Kim, Hyung-Jin Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.7

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 > $1.0{\times}10^7$.

KN-12 운반용기를 이용한 고리 사용후핵연료 소내수송.저장

정성환,백창열,최병일,양계형,이대기,Chung, Sung-Hwan,Baeg, Chang-Yeal,Choi, Byung-Il,Yang, Ke-Hyung,Lee, Dae-Ki 한국방사성폐기물학회 2006 방사성폐기물학회지 Vol.4 No.1

Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA's transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.