A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

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.

International Atomic Energy Agency defines the term “Poison” as a substance used to reduce reactivity, by virtue of its high neutron absorption cross-section, in IAEA glossary. Poison material is generally used in the reactor core, but it is also used in dry storage systems to maintain the subcriticality of spent fuel. Most neutron poison materials for dry storage systems are boron-based materials such as Al-B Carbide Cermet (e.g., Boral®), Al-B Carbide MMC (e.g., METAMIC), Borated Stainless Steel, Borated Al alloy. These materials help maintain subcriticality as a part of the basket. U.S.NRC report NUREG-2214 provides a general assessment of aging mechanisms that may impair the ability of SSCs of dry storage systems to perform their safety functions during longterm storage periods. Boron depletion is an aging mechanism of neutron poison evaluated in that report. Although that report concludes that boron depletion is not considered to be a credible aging mechanism, the report says analysis of boron depletion is needed in original design bases for providing long-term safety of DSS. Therefore, this study aimed to simulate the composition change of neutron poison material in the KORAD-21 system during cooling time considering spent fuel that can be stored. The neutron source term of spent fuel was calculated by ORIGEN-ARP. Using that source term, neutron transport calculation for counting neutrons that reach neutron poison material was carried out by MCNP®-6.2. Then, the composition change of neutron poison material by neutron-induced reaction was simulated by FISPACT-II. The boron-10 concentration change of neutron poison material was analyzed at the end. This study is expected to be the preliminary study for the aging analysis of neutron poison material about boron depletion.

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.

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

원자력발전소내 습식저장중인 사용후핵연료의 건식저장을 위해서는 캐니스터 내부에 사용후핵연료를 옮겨 담은 이후, 건 식저장 캐니스터 내장품이나 사용후핵연료 다발의 부식방지 및 피복관의 열화방지를 위해 모든 수분은 제거해야 한다. 캐 니스터 내부의 수분을 제거하는데 사용할 수 있는 기체강제순환 건조시스템 개발을 위한 연구개발이 진행중이다. 본 연구 에서는 기체강제순환 건조시스템 설계, 제작을 위한 예비설계를 수행하였다. 예비설계에는 캐니스터 내부 잔존수분 제거를 위한 건조사례조사, 건조관련 규격이나 표준, 건조합격기준, 건조장치구성, 현장요구분석, 습식저장중인 사용후핵연료 특성 을 포함하였다. 예비설계를 통하여 기체강제순환 건조시스템의 설계 개념도와 P&ID를 도출하였고, 이를 활용하여 건조시 스템 제작을 위한 상세설계를 수행할 것이다.

국가 주도로 2009년부터 개발중인 경수로 사용후핵연료 건식저장시스템은 금속 겸용용기와 콘크리트 저장용기의 두 가지 방식이다. 국외 건식저장시설 운영 시 주요 격납감시 방안으로는 금속 겸용용기인 경우 이중 뚜껑 사이에 압력센서를 설치 하여 실시간 압력변화를 감시하는 방법이 있고 콘크리트 저장용기의 경우는 캐니스터 기반으로 주요 격납 경계인 뚜껑을 이중으로 용접하는 방식으로 구조물(over pack 또는 module)의 공기 유로인 입구 및 출구에 대한 온도 변화를 감시하는 방 법으로 격납을 관리하는 것으로 나타났다. 미국, 독일 등 30 년 이상 안정적으로 저장시설을 관리한 국가의 다양한 적용기 술 및 운영사례를 조사/분석하여 우리가 개발중인 저장시스템에 적용할 수 있는 격납감시 방안을 도출하는데 활용할 수 있 도록 하였다.