The objective of this study was to compare the feeding effects of imported timothy hay, domestic Italian ryegrass silage (IRGS) and IRG dried by hot-air dry system (IRGHDS) under basic total mixed ration (bTMR) on rumination activity, milk production, and milk composition in lactating dairy cows. Eighteen Holstein dairy cows were divided into three groups: control (bTMR + imported timothy hay), treatment 1 (bTMR + IRGS) and treatment 2 (TMR + IRGHDS) groups. The study was conducted over a total period of 24 days, including a 10 days adaptation period and a 14 days main experimental period. The results indicate body weight was not significantly different between the three groups (p>0.05), and rumination time was significantly higher in the treatment 1 group compared with the control group (p<0.001). Milk fat, protein, and lactose were not significantly different between the three groups (p>0.05), however, somatic cell counts were significantly lower in the treatment 1 group compared with the control group (p<0.001). As a result of the economic feasibility analysis, it is expected that feeding TMR using IRGS and IRGHDS will increase profit by 402.8 won/day and 331.4 won/day per cow compared to imported timothy hay. Therefore, IRGS and IRGHDS can be used as substitutes for imported timothy hay in feeding lactating dairy cows.

배관 세정 공정은 조선소 선박 건조과정에서 배관 설치 후 장비를 시운전하는 단계로 넘어가기 전에 최종적으로 배관 내부의 이물질을 제거하는 매우 중요한 공정이다. 만약에 배관 내에 이물질이 있는 상태에서 장비를 시운전하는 단계로 넘어갈 경우 이물질이 고가의 장비에 유입되어 펌프 및 기어, 베어링 등이 파손되는 요인이 된다. 특히 펌프나 유압 밸브 같은 경우는 아주 작은 이물질이라도 장비 속으로 유입이 되면 대형 사고로 이어지며, 이런 장비 사고는 주변에 장비 운전을 하는 작업자의 인명사고까지 연계되어 중대 재해 의 잠재 원인이 되므로 조선소 고객인 선주들도 매우 집중적으로 확인하고 관리하는 공정이다. 이러한 문제점을 해결하기 위해 본 연구 에서 기존의 배관 세정 공법에서 유세정 효과를 증가시킬 수 있도록 배관 내 세정 유체의 흐름을 증가할 수 있는 시스템을 제안한다.

The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

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.

This paper describes the development and operation of an autonomous robotic system designed for pyroprocess automation. The unique challenges of pyroprocess automation, such as the need for a highly dry atmosphere to handle materials like chloride, are addressed through this system. For the experiments, a specialized dehumidifier and dry mock-up facility were designed to produce dry air condition. Performances of dry air conditioning for the various simulated situations were evaluated, including assessing worker access within a mock-up to determine the system’s feasibility. To enable automation, containers used for processing materials were modified to fit the gripper system of the gantry robot. The loading and unloading of materials in each equipment were automated to connect them with the robotic system. This gantry robot primarily utilized macro motions to approach waypoints containing process materials, reducing the need for precise approach motions. Its tapered jaw design allowed it to grip target objects even with imperfect positioning. The robot’s motions were programmed initially using a robot simulator for positioning and motion planning, and real-world accuracy was tested in a dry mock-up facility using the OPC platform. Finally, the paper discusses the potential application of XR (eXtended Reality) technology in this context, which could enhance the robot’s operation and provide valuable insights into the automation process. Further analysis of XR technology’s feasibility and benefits for this specific pyroprocess automation system are presented.

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.

As the amount of on-site Spent Nuclear Fuel (SNF) in storage increases due to the continued operation of Nuclear Power Plants (NPPs) in Korea, the on-site wet storage pool is expected to become saturated. Therefore, a facility for safely storing the spent nuclear fuel is required so that there is no problem with operation of the NPP until permanent disposal of SNF. Prior to the construction of such a facility, the safety analysis of the interim storage facility and verification of the safety of the spent fuel storage system (e.g. cask, silo) to be used are required according to Article 63 of the Nuclear Safety Act. In this process, analysis of the Structures, Systems, and Components (SSCs) of the storage system is needed. Based on the analysis, it is necessary to efficiently classify SSCs that are important to safety in order to differentiate management that more thoroughly manages those important to safety. In Korea, according to the notice of the Nuclear Safety and Security Commission, the components performing essential safety functions for the safe storage of spent fuel storage system are to be classified as “important safety equipment”. 10 CFR Part 72, a federal regulation related to interim storage facilities in the United States, also requires the identification of SSCs that fall under “Important to Safety (ITS)”, which is like domestic case. In addition, it has been confirmed that there are cases in which detailed classification according to Reg Guide 7.10 and NUREG-CR/6407 is added in Safety Analysis Report. However, these existing classification methods are not only classified as a single grade except for the method according to the Reg guide, but all are classified according to a qualitative standard. Qualitative criteria may cause ambiguity in judgment, resulting in subjective judgment of the person who proceeds in the classification process. Therefore, in this study, a new classification method is proposed to solve the problem according to the qualitative classification method. Assessing the level of radiological harm to the general public due to the assumption of failure of SSC in the spent fuel storage system is used as a quantitative evaluation standard.

For Dry Storage of Spent Nuclear Fuel (SNF), all moisture must be removed from the dry storage canister through subjected to a drying process to ensure the long-term integrity. In NUREG-1536, the evacuation of most water contained within the canister is recommended a pressure of 0.4 kPa (3 torr) to be held in the canister for at least 30 minutes while isolated from active vacuum pumping as a measure of sufficient dryness in the canister. In the existing drying process, the determination of drying end point was determined using a dew point sensor indirectly. Various methods are being studied to quantify the moisture content remaining inside the canister. We presented a moisture quantification method using the drying process variables, like as temperature, pressure, and relative humidity operation data. During the drying process, it exists in the form of a mixed gas of water vapor and air inside the canister. At this time, if the density of water vapor in the mixed gas discharged out of the canister by the vacuum pump is known, the mass of water removed by vacuum drying can be calculated. The canister is equipped with a pressure gauge, thermometer and dew point sensor. The density of water vapor is calculated using the pressure, temperature and relative humidity of the gas obtained from these sensors. First, calculate the saturated water vapor pressure, and then calculate the humidity ratio. The humidity ratio refers to the ratio of water vapor mass to the dry air mass. After calculating the density of dry gas, multiply the density by the humidity ratio to calculate the density of water vapor (kg/m3). Multiply the water vapor density by the volume flow (m3/s) to obtain the mass value of water (kg). The calculated mass value is the mass value obtained per second since it is calculated through the flow data obtained every second, and the amount of water removed can be obtained by summing all the mass values. By comparing this value with the initial moisture content, the amount of moisture remaining inside the canister can be estimated. The validity of the calculations will be verified through an experimental test in the near future. We plan to conduct various research and development to quantify residual water, which is important to ensure the safety of the drying process for dry storage.

To dry storage of spent nuclear fuel withdrawn the wet storage, all moisture inside the dry storage container must be removed to ensure the long-term integrity and retrievability. Substantial amounts of residual water in dry storage container may have potential impacts on the fuel, cladding, and other components in the dry storage system, such as fuel degradation and cladding corrosion, embrittlement, and breaching. The drying could perform as a vacuum drying process or a forced helium dehydration process. In NUREG-1536, the evacuation of most water contained within the canister is recommended a pressure of 0.4 kPa (3 torr) to be held in the canister for at least 30 minutes while isolated from active vacuum pumping as a measure of sufficient dryness in the canister. Monitoring the moisture content in gas removed from the canister is considered as a means of evaluating adequate dryness. Dew point monitoring and special techniques could be used to evaluate this adequacy. Various studies are continuing for quantitative evaluation of residual moisture inside the dry storage system. Andrawes proposed a methodology for determining trace water contents in gaseous mixtures, utilizing gas chromatography together with a helium ionization source. A microwave plasma source and emission spectrometry were utilized to determine trace amounts of bound water in solid samples using peak areas of atomic oxygen (O) and hydrogen (H) emissions. Bryans measured the gas samples taken from the High Burn-Up Demonstration Cask at three intervals: 5 hours, 5 days, and 12 days after the completion of drying and backfilling in the North Anna power Station. To measure water content, a Vaisala humidity probe was used. Final results indicated that the cask gas water content built up over 12 days to a value of 17,400 ppmv ±10%, equivalent to approximately 100 g of water within the entire cask gas phase. Tahiyats also proposed a methodology that involves a direct current (dc) driven plasma discharge and optical emission spectroscopy for detecting and quantifying water vapor in a flowing gas stream under both trace and high water vapor loading conditions. For detecting water vapor concentration, the emission from H at 656.2 nm was employed. The H emission is the red visible spectral line generated by a hydrogen atom when an electron falls from the third lowest to the second lowest energy level, this suggests that the normalized H intensity can be used as a marker for water vapor detection and quantification. Several of the attempts are continuing to quantify water contents in dry storage system. Lessons learned by Case studies would be provided insights into how to improve future measurements.

Concrete structures of spent nuclear fuel interim storage facility should maintain their shielding ability and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain its unique functions (shielding and structural integrity). Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In accident conditions such as tip over and aircraft collision, both static material properties and dynamic properties of the concrete are required to evaluate the structural integrity of the concrete structures. Unlike the calculated damage results for the static deformation of the concrete structure, it is very difficult to accurately estimate the damage values of the degraded concrete structures where an aircraft collides at a high strain rate. Therefore, the present authors have a plan to establish a database of the dynamic material properties of deteriorated concrete and implement to a Finite Element Analysis model. Prior to that, dynamic increase factors described in a few technical specifications were investigated. The dynamic increase factor represents the ratio of the dynamic to static strength and is normally reported as function of strain rate. In ACI-349, only the strain rate is used as a variable in the empirical formula obtained from the test results of specified concrete strengths of 28 to 42 MPa. The maximum value of dynamic increase factor is limited to 1.25 in the axial direction and 1.10 in the shear direction. On the other hand, in the case of the CEB model, static strength is included as variables in addition to the strain rate, and a constitutive equation in which the slope changes from the strain rate of 30 /s is proposed. As plotting the two dynamic increase factor models, in the case of ACI, it is drawn as a single line, but in the case of CEB, it is plotted as multiple lines depending on the static strength. The test methods and specimen sizes of the previously performed tests, which measured the concrete dynamic properties, were also investigated. When the strain rate is less than 10 /s, hydraulic or drop hammer machines were generally used and the length of the specimens was more than twice the diameter in most cases. However, in the case of Split Hopkinson Pressure Bar tests, the small size specimens are preferred to minimize the inertia effect, so the specimens were small and the length was less than twice the diameter. We will construct the dynamic properties DB with our planned deteriorate concrete specimen test, and also include the dynamic property data already built in the previous studies.