It is very important that the confinement of a spent fuel storage systems is maintained because if the confinement is damaged, the gaseous radioactive material inside the storage cask can leak out and have a radiological impact on the surrounding public. For this reason, leakage rate tests using helium are required for certificate of compliance (CoC) and fabrication inspections of spent fuel storage cask. For transport cask, the allowable leakage rate can be calculated according to the standardized scenario presented by the IAEA. However, for storage cask, the allowable leakage rate is determined by the canister, facility, and site specific information, so it is difficult to establish a standardized leakage rate criterion. Therefore, this study aims to establish a system that can derive system-specific leakage test criteria that can be used for leakage test of actual storage systems. First, the variables that can affect the allowable leakage rate for normal and accident conditions were derived. Unlike transportation systems, for storage systems, the dose from the shielding analysis and the dose from the confinement analysis are summed up to determine whether the dose standard is satisfied, and even the dose from the existing nuclear facilities is summed up during normal operation condition. For this reason, the target dose is used as an input variable when calculating the allowable leakage rate for the storage system. In addition, the main variables are the distance from the boundary of the exclusive area, the number of cask, the inventory of nuclide material in the cask, the free volume, and the internal and external pressure. Utilizing domestic and US NRC guidelines, we derived basic recommended values for the selected variables. The GASPARII computer code that can evaluate the dose to the public under normal operating conditions was utilized. Using the above variables, the allowable leakage rate is calculated and converted to the allowable criteria for helium leakage rate test. The developed system was used to calculate the allowable leakage rate for normal and accident conditions for a hypothetical storage system. The leakage rate criteria calculation system developed in this study can be useful for CoC and fabrication inspections of storage systems in the future, and a GUI-based program will be built for user convenience.
Considering the domestic situation where all nuclear power plants are located on seaside, the interim storage site is also likely to be located on coastal site. Maritime transportation is inevitable and the its risk assessment is very important for safety. Currently, there is no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the immersion of transport cask. Previous studies show that the release rate of radionuclides contained in a submerged transport cask is significantly affected by the area of flow path generated at the breached containment boundary. Due to the robustness of a cask, the breach is the most likely generated between the lid and body of cask. CRIEPI investigated the effect of cask containment on the release rate of radioactive contents into the ocean and proposed a procedure to calculate the release rate considering the socalled barrier effect. However, the contribution of O-ring on the release rate was not considered in the work. In this study, test and analysis is performed to determine the equivalent flow path gap considering the influence of O-rings. These results will be implemented in the computational model to assess sea water flow through a breached containment boundary using CFD techniques to assess radionuclide release rates. To evaluate the release rate as a function of lid displacement, a small containment vessel is engineered and a metal O-ring of the Helicoflex HN type is installed, which is the most commonly used one in transport and storage casks. The lid of containment vessel is displaced in vertical and horizontal direction and the release rate of the vessel was quantified using the helium leak test and the pressure drop test. Through this work, the relationship between the vertical opening displacement and horizontal sliding displacement of the cask lid and the actual flow path area created is established. This will be implemented in the CFD model for flow rate calculation from a submerged transport cask in the deep sea. In addition, the compression of the O-ring causes very small gaps, such as capillaries. In these cases, Poiseuille’s law is used to calculate the capillary flow rate.
Gate valves are hydraulic components used to shut-off the water flow in water distribution systems. Gate valves may fail owing to various aspects such as leakage through seats, wearing of packing, and corrosion. Because it is considerably challenging to detect valve malfunctioning until the operator identifies a significant fault, failure of the gate valve may lead to a severe accident event associated with water distribution systems. In this study, we proposed a methodology to diagnose the faults of gate valves. To measure the pressure difference across a gate valve, two pressure transducers were installed before and after the gate valve in a pilot-scaled water distribution system. The obtained time-series pressure difference data were analyzed using a machine learning algorithm to diagnose faults. The validation of whether the flow rate of the pipeline can be predicted based on the pressure difference between the upstream and downstream sides of the valve was also performed.
Considering the domestic situation where all nuclear power plants are located on seaside, the interim storage site is also likely to be located on coastal site. Maritime transportation is inevitable and the its risk assessment is very important for safety. Currently, there is no independently developed maritime transportation risk assessment code in Korea, and no research has been conducted to evaluate the release of radioactive waste due to the immersion of transport cask. Previous studies show that the release rate of radionuclides contained in a submerged transport cask is significantly affected by the area of flow path generated at the breached containment boundary. Due to the robustness of a cask, the breach is the most likely generated between the lid and body of cask. CRIEPI investigated the effect of cask containment on the release rate of radioactive contents into the ocean and proposed a procedure to calculate the release rate considering the so-called barrier effect. However, the contribution of O-ring on the release rate was not considered in the work. In this study, test and analysis is performed to determine the equivalent flow path gap considering the influence of O-rings. These results will be implemented in the computational model to assess sea water flow through a breached containment boundary using CFD techniques to assess radionuclide release rates. The evaluation of release rate due to container lid gaps has been performed by CRIEPI and BAM. In CRIEPI, the gap of the flow path was calculated from the roughness of the container surface without a quantitative assessment of the severity of the accident. In this work, to evaluate the release rate as a function of lid displacement, a small containment vessel is engineered and a metal Oring of the Helicoflex HN type is installed, which is the most commonly used one in transport and storage casks. The lid of containment vessel is displaced in vertical and horizontal direction and the release rate of the vessel was quantified using the helium leak test and the pressure drop test. Through this work, the relationship between the vertical opening displacement and horizontal sliding displacement of the cask lid and the actual flow path area created is established. This will be implemented in the CFD model for flow rate calculation from a submerged transport cask in the deep sea.
Waterworks facilities inevitably experience some amount of leakage even if there is a lot of investment or state-of-the-art technology that is applied such as DMA(District Metered Area) system construction, leakage detection, repair, pipe rehabilitation, etc. The primary reason is the leakage is naturally restored over time. In the UK, this restoration characteristic is defined as NRR(Natural rate of rise of leakage) and used to decision making for prioritizing active leakage control of DMAs. However, this restoration characteristic is well recognized, but researches on NRR in the water distribution system are insufficient in Korea. In this study, the estimation method of NRR was developed suitable for applicating in Korea considering of SCADA data, water infrastructure, and water usage patterns by modification of the UK's NRR method. The proposed method was applied to 9 DMAs and verified it's applicability by comparing with the other water loss performance indicators. It is expected that the proposed method can be used to support decision making for sustainable NRW(Nor-revenue water) management in the water distribution system.
변화하는 환경조건에서 종자의 장기간 저장시발아율, 용출정도 및 수분흡수율을 밝혀서 장기저장 가능한 품종의 선발기준과 그의 특성을 알아보기 위하여 본 실험을 실시하였던 바, 다음과 같은 결과를 얻었다. 1. 콩나물재배기와 petri dish에서의 발아율은 20개월 저장시 품종간의 차이가 컸으며, 8개월 저장에서도 차이가 있었다. 2. 배축의 길이와 두께는 저장기간 간에 유의한 차이가 있었다. 3. 용출정도는 20개월 저장에서 팔달콩과 풀무원이 SS88038계통보다 2배 정도 높았다. 4. 초기 30분의 수분흡수율은 20개월 저장에서는 품종간 차이가 있었으며, 8개월 저장에서는 팔달콩과 풀무원이 SS88038보다 1.5배와 2배 빨랐다. 5. 발아율과 용출정도는 20개월 저장에서 고도로 유의한 부의 상관을 나타내었고, 100립중은 발아율 및 용출정도와 상관이 없었다. 6. 수분흡수율과 용출정도는 8개월 저장에서 침종 30분과 1시간에서 고도로 유의한 정의 상관을 나타내었고, 20개월 저장에서는 침종 30분, 1시간, 3시간에서 고도로 유의한 정의 상관을 나타내었으며, 주공의 열림과 수분흡술율은 상관이 없었다.