Our study examined a total of 60 dead wood samples (Quercus spp.), collecting 30 samples each in summer and winter, and measured various environmental variables in the field. These samples were classified into three decay classes: fresh, intermediate, well-decayed. We sampled saproxylic beetles within the collected dead wood using emergence traps. Conducted a cluster analysis to explore their relationships of saproxylic beetle and environmental variables. Our results revealed that season and decay class were significant environmental variables affecting saproxylic beetle communities. These results highlight the sensitivity of saproxylic beetles to fluctuations of environment and climate. In summary, our study highlights the intricate relationships between environmental variables and saproxylic beetles and suggests that different types of dead wood should be maintained in forests.

In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM−1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

The thermal evaluations for the conceptual design of the deep geological repository considering the improved modeling of the spent fuel decay heat were conducted using COMSOL Multiphysics computational program. The maximum temperature at the surface of a disposal canister for the technical design requirement should not exceed 100°C. However, the peak temperature at the canister surface should not exceed 95°C considering the safety margin of 5°C due to several uncertainties. All thermal evaluations were based on the time-dependent simulation from the emplacement time of the canister to 100,000 years later. In particular, the heat source condition was set to the decay heat rate and axial decay heat profile of the PLUS7 fuel with 4.0wt% U-235 and 45 GWD/MTU. The thermal properties of the granitic rock in South Korea were applied to the host rock region. For the reference design case, the cooling time of the SNF was set to 40 years, the distance between the deposition holes 8 meters and that between the deposition tunnels 30 meters. However, the peak temperature at the canister surface at 10 years was 95.979°C greater than 95°C. This design did not meet the thermal safety requirement and needed to be modified. For the first modified case, when the distance between the deposition tunnels was set to 30 meters, three cooling time cases of 40, 50 and 60 years and five distances of 6, 7, 8, 9 and 10 meters between the deposition holes were considered. The design with the distances of 9 and 10 meters between the deposition holes for the cooling time of 40 years and all five distances for 50 and 60 years were less than 95°C. For the second modified case, when the distance between the deposition holes was set to 8 meters, three cooling time cases of 40, 50 and 60 years and five distances of 20, 25, 30, 35 and 40 meters between the deposition tunnels were considered. The design with the distances of 35 and 40 meters between the deposition tunnels for the cooling time of 40 years, the distances of 25, 30, 35 and 40 meters for 50 years and all five distances for 60 years were less than 95°C. As a result, the peak temperature at the canister surface decreased as the cooling time and the distance between the deposition holes and the tunnels increased.

In the 3rd revision of NUREG-0800, which was revised in 2007, the calculation method for decay heat in the design of the Ultimate Heat Sink (UHS) for a pressurized water reactor is recommended to be based on the ANSI/ANS-5.1 method. This method employs a more complex decay heat calculation formula compared to the one introduced in Branch Technical Position ASB 9-2, which was presented in the 2nd revision. While most of the variables for decay heat calculation in ANSI/ANS-5.1 can be inferred from the methods outlined in the appendices, determining the fractions of fission products is not straightforward despite their significant impact on the results. When reviewing documents that evaluate decay heat using the ANSI/ANS-5.1 method, it is observed that they often adopt a conservative approach by assuming that the fraction of the most influential fission product is 100%. In this study, the fractions of each fission product presented in LLNL’s 2016 report were used to calculate decay heat, and the results were compared with the ASB 9-2 method and ORIGEN code results. The comparison showed that ANS 5.1 tends to yield higher decay heat values than ANS 9-2, particularly at the reference time of 1M seconds, while ORIGEN-ARP generally produced lower values. Therefore, it is concluded that even when using the ANSI/ANS-5.1 method with the fractions of each fission product for decay heat calculations in spent nuclear fuel wet or dry storage facility assessments, it provides a sufficiently conservative thermal evaluation.

본 연구에서는 어선의 운동성능을 향상하기 위해 부착되는 부가물의 조합과 파라미터 변경에 따른 어선의 자유 횡 동요 감쇠 와 저항 성능을 평가하였다. 성능 평가를 위해 전산유체역학(Computational Fluid Dynamics)을 이용한 수치해석을 수행하였으며, 주요 부가 물인 빌지킬과 선저킬의 조합과 치수 변경에 따른 횡 동요 주기와 감쇠 계수의 변화를 확인하였다. 선저킬의 경우 길이가 변화함에 따른 횡 동요 감쇠 계수의 변화가 상대적으로 크지 않음을 확인하였다. 반면 빌지킬의 경우 길이와 각도의 증가에 따라서 횡 동요 감쇠 계수가 증가함을 확인하였다. 4가지 부가물 조합 조건과 나선의 저항 성능을 비교하였으며, 부가물에 의한 어선의 자세와 압력분포의 변화로 인 해 저항이 증가함을 확인하였다. 본 연구 결과를 통해 부가물 크기와 배치가 어선의 운동 및 저항 성능에 미치는 영향을 확인할 수 있었 으며, 어선 적용 시에 도움이 될 수 있을 것으로 기대한다.

Currently, the most widely accepted disposal concept for long-term isolation of high level radioactive waste including spent nuclear fuels is to disposal in a deep geological repository designed and constructed with multiple barriers composed of engineered and natural barriers so that the waste can be completely isolated in a stable deep geological environment. In this concept, an important consideration is the heat generated from the waste due to the large amount of fission products present in the high level waste loaded in the disposal container. For safe and complete isolation of high level radioactive waste in the deep geology, the disposal concepts that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. In this paper, the deposition hole configuration and the decay heat dissipation area (surface area) of disposal container were considered for the efficient thermal management in the deep geological disposal concept. Heat transfer through the waste form, its container and surrounding components and the rock will be mainly by conduction. Heat transfer by radiation and convection can be negligible after backfilling. When considering heat conduction, according to Fourier’s law, if the thermal conductivity of the repository components is the same, the greater the heat dissipation area and the adjacent temperature gradient, the greater the conduction effect. Therefore, rather than the conventional concept of loading 4 PWR spent fuel assemblies per disposal container and placing one disposal container in a deposition hole, it is better to load one assembly per disposal container and place 4 disposal containers in a deposition hole. In this case, it was found that the disposal area could be reduced through efficient thermal management. Considering this thermal management method as an alternative to the concept of deep geological disposal, additional research is needed.

This paper mainly focuses on the maximum decay heat estimation generated from spent fuel assemblies in the spent fuel pool of Kori units 3&4 at the beginning decommissioning. It is assumed that the spent fuel pool is fully occupied with 2,260 spent fuel assemblies, same as its design capacity. In addition, equally 56.5 spent fuel assemblies have been generated per year. The minimum cooling time is five years considering the transition phase between the permanent shutdown and the amendment of Operating License for decommissioning. Sending and receiving of spent fuel assemblies to/from other units are neglected. Seven representative spent fuel assembly groups are established based on the burnup rate and cooling time. Conservatively high values for the burnup rates and low values for the cooling times are applied. Calculation of the decay heat of each representative group has been performed by using ORIGEN decay solver of SCALE. Then, total decay heat has been calculated based on this. Group 1, 2, and 3 contain comparatively old spent fuel assemblies with 45 GWd/tU burnup rate and 20~30 cooling years. The calculation shows 489~586 watts of decay heat per assembly. Group 4, 5, 6, and 7 contain comparatively new spent fuel assemblies with 55 GWd/tU burnup rate and 5~20 cooling years. The calculation shows 741~1,483 watts of decay heat per assembly. The total maximum decay heat therefore is estimated as 1,609,459 watts.

The research for the safe management of high-level waste in Korea has been conducted by the Korea Atomic Energy Research Institute since 1997, and the results have formed the basis of the national basic plan for the high-level waste management and the revised national basic plan. In the future, it is evolving and developing R&D focusing on securing technologies for demonstration of the disposal technologies and R&D to develop disposal concepts that increase safety and improve efficiency. Efficient management of heat generated from high-level radioactive waste, including spent nuclear fuel, is an important factor in establishing the disposal concepts because it must be in harmony with key factors such as repository layout, waste disposal container specifications, and design and operation for the barriers of the disposal system. For safe and complete isolation of highlevel radioactive waste in the deep geology, the disposal systems that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. These disposal systems were based on low burn-up spent nuclear fuel characteristics generated in the early stages of nuclear power generation, and next, based on the high-level wastes from recycling process of the high burn-up spent nuclear fuels, and were the direct disposal systems for the high burn-up spent nuclear fuels. So, it is necessary to track and analyze the change process in the decay heat characteristics of the high-level waste to be disposed of in order to improve the disposal concept, which enhances the safety of disposal and the utilization of the national land. Therefore, in this paper, the process of change in decay heat of reference spent nuclear fuels for disposal applied to the disposal concepts from the initial stage of development of high-level waste disposal technology to the present in Korea is analyzed.

공주지역에 조림된 리기다소나무 군락에서 국내 대표 수종인 소나무와 방풍림으로 주로 조림되는 곰솔 낙엽의 분해율 및 분해과정에 따른 영양염류의 함량 변화를 파악하였다. 분해 60개월 경과 후 소나무 낙엽과 곰솔 낙엽의 잔존율은 각각 42.12±5.30과 24.79±1.98%로 소나무와 곰솔의 낙엽 분해율은 곰솔 낙엽의 분해가 소나무 낙엽의 분해에 비해 빠르게 일어났다. 60개월 경과 후 소나무 낙엽과 곰솔 낙엽의 분해상수 (k)는 각각 3.02과 3.59로 곰솔 낙엽의 분해상수가 다소 높게 나타났다. 소나무 낙엽의 분해과정에 따른 C/N, C/P 비율은 초기에 각각 14.4, 144.1 이었으나 60개월 경과 후에는 각각 2.26와 40.1로 점차 감소하였으며, 곰솔 낙엽의 경우 초기 C/N, C/P 비율은 각각 14.4와 111.3로 나타났고, 60개월 경과 후에는 각각 3.06와 45.8로 나타났다. 낙엽의 초기 N, P, K, Ca, Mg 함량은 소나무 낙엽에서 각각 3.07, 0.31, 1.51, 16.56, 2.03 mg g-1, 곰솔 낙엽에서 각각 3.02, 0.39, 0.99, 19.55, 1.48 mg g-1로 소나무 낙엽과 곰솔 낙엽의 질소와 인의 함량은 유사하였다. 60 개월 경과 후 N, P, K, Ca, Mg의 잔존율은 소나무 낙엽에서 각각 231.08, 130.13, 35.68, 48.58, 36.03%이었고, 곰솔 낙엽에서 각각 143.91, 74.02, 28.59, 45.08, 44.99%로 나타났다.

We present numerical simulations of decaying hydrodynamic turbulence initially driven by solenoidal (divergence-free) and compressive (curl-free) drivings. Most previous numerical studies for decaying turbulence assume an isothermal equation of state (EOS). Here we use a polytropic EOS, P ∝ ργ, with polytropic exponent γ ranging from 0.7 to 5/3. We mainly aim at determining the effects of γ and driving schemes on the decay law of turbulence energy, E ∝ t-α. We additionally study probability density function (PDF) of gas density and skewness of the distribution in polytropic turbulence driven by compressive driving. Our findings are as follows. First of all, we find that even if γ does not strongly change the decay law, the driving schemes weakly change the relation; in our all simulations, turbulence decays with α ≈ 1, but compressive driving yields smaller α than solenoidal driving at the same sonic Mach number. Second, we calculate compressive and solenoidal velocity components separately and compare their decay rates in turbulence initially driven by compressive driving. We find that the former decays much faster so that it ends up having a smaller fraction than the latter. Third, the density PDF of compressively driven turbulence with γ > 1 deviates from log-normal distribution: it has a power-law tail at low density as in the case of solenoidally driven turbulence. However, as it decays, the density PDF becomes approximately log-normal. We discuss why decay rates of compressive and solenoidal velocity components are different in compressively driven turbulence and astrophysical implication of our findings.

This paper gives two graph-based algorithms for radioactive decay computation. The first algorithm identifies the connected components of the graph induced from the given radioactive decay dynamics to reduce the size of the problem. The solutions are derived over the precalculated connected components, respectively and independently. The second algorithm utilizes acyclic structure of radioactive decay dynamics. The algorithm evaluates the reachable vertices of the induced system graph from the initially activated vertices and finds the minimal set of starting vertices populating the entire reachable vertices. Then, the decay calculations are performed over the reachable vertices from the identified minimal starting vertices, respectively, with the partitioned initial value over the reachable vertices. Formal arguments are given to show that the proposed graph inspired divide and conquer calculation methods perform the intended radioactive decay calculation. Empirical efforts comparing the proposed radioactive decay calculation algorithms are presented.

This study developed prediction models of chlorine bulk decay coefficient by each condition of water quality, measuring chlorine bulk decay coefficients of the water and water quality by water purification processes. The second-reaction order of chlorine were selected as the optimal reaction order of research area because the decay of chlorine was best represented. Chlorine bulk decay coefficients of the water in conventional processes, advanced processes before rechlorination was respectively 5.9072 (mg/L)-1d-1 and 3.3974 (mg/L)-1d-1, and 1.2522 (mg/L)-1d-1 and 1.1998 (mg/L)-1d-1 after rechlorination. As a result, the reduction of organic material concentration during the retention time has greatly changed the chlorine bulk decay coefficient. All the coefficients of determination were higher than 0.8 in the developed models of the chlorine bulk decay coefficient, considering the drawn chlorine bulk decay coefficient and several parameters of water quality and statistically significant. Thus, it was judged that models that could express the actual values, properly were developed. In the meantime, the chlorine bulk decay coefficient was in proportion to the initial residual chlorine concentration and the concentration of rechlorination; however, it may greatly vary depending on rechlorination. Thus, it is judged that it is necessary to set a plan for the management of residual chlorine concentration after experimentally assessing this change, utilizing the methodology proposed in this study in the actual fields. The prediction models in this study would simulate the reduction of residual chlorine concentration according to the conditions of the operation of water purification plants and the introduction of rechlorination facilities, more reasonably considering water purification process and the time of chlorination. In addition, utilizing the prediction models, the reduction of residual chlorine concentration in the supply areas can be predicted, and it is judged that this can be utilized in setting plans for the management of residual chlorine concentration.