In order to determine fragility curves, the limit state of piers for each damage level is suggested in this paper based on the previous test results in Korea, including our test results. In previous studies, the quantitative measures for damage levels of piers have been represented by curvature ductility, lateral drift ratio, or displacement ductility. These measures are transformed to lateral drift ratios of piers for consistency, and the transformed values are compared and verified with our push-over test results for flexural RC piers with a circular cross-section. The test specimens are categorized concerning the number of lap-splices in the plastic hinge region and whether seismic design codes are satisfied or not. Based on the collected test results in Korea, including ours, the lateral drift ratio for each pier damage level is suggested.

해상공사에서 발생하는 부유사는 해수의 탁도를 증가시키고 광량을 감소시켜 해양생물에 악영향을 미치므로 해양환경영향평 가에서 중요한 요소이다. 하지만 평가에 적용되는 인자에 대한 공식적인 자료의 부족과 평가자의 능력에 따라 그 영향이 달리 평가되고 있다. 따라서 본 연구에서는 해역이용영향평가센터에서 검토한 3년간(2012–2014)의 매립, 준설, 외곽시설물 설치 등 총 58건 사업에 대한 부유사 확산 평가에 대한 실태를 진단하고 개선방안을 제시하였다. 개선방안 제시를 위해 4가지의 평가지표(격자체계의 적정성, 원단위의 적정성, 대표입경 및 침강속도의 적정성)를 적용하였다. 각 항목별 신뢰도에 평균점수 분석결과, 격자체계는 25점, 원단위는 60점, 대표입 경은 34점 그리고 침강속도는 17점으로 평가항목에 대한 개선방안이 필요한 것으로 나타났다. 본 연구에서는 부유사 확산 평가상태에 대 한 진단 및 신뢰도 평가 결과를 활용하여 부유사 확산예측에 대한 개선방안을 제안하였다. 먼저, 부유사 발생원단위 및 대표입경별 침강 속도에 대한 공신력 있는 값이 가이드라인을 통해 제공해야 한다. 그리고 실무에선 신뢰성 향상을 위해 격자체계의 적정성과 결과의 검 증을 철저히 해야 한다.

토마토의 생육과 수확량을 예측하기 위한 중요한 요소 중의 하나는 엽면적이다. 이러한 엽면적을 정확하게 예측하는 것 은 토마토 식물 생장 평가 모델의 시작이라고 할 수 있다. 이를 위해 본 연구는 토마토 잎의 측정을 통해 엽면적(LA)을 추정 하는 효과적인 모델을 확인하기 위해 수행하였다. 토마토 식 물 잎 조사를 위해 2주 간격으로 5개체의 토마토 식물체의 전 개된 모든 잎에 대해 엽면적(LA), 엽장(L), 엽폭(W), 엽신장 (La)를 측정하였다. LA와 토마토 잎 독립변수의 상관관계는 La × W, L × W, La + W, L + W의 식이 강한 양의 관계를 나타 냈다. LA 추정은 LA = a + b(La2 + W2) 을 사용하는 선형 모델 이 가장 정확한 추정치를 나타내었다(R2 = 0.867, RMSE = 88.76). 9월부터 12월까지 토마토 잎의 위치에 따른 상, 중, 하 엽의 모델을 살펴본 결과, 상, 중, 하로 잎 위치에 따른 모델별 결정계수(R2) 값은 각각 0.878, 0.726, 0.794였다. 상위엽을 바탕으로 추정된 모델의 정확도가 가장 높았는데, 이는 10월 이후 토마토 재배 농가에서 중위엽과 하위엽에 실시한 반적엽 의 영향으로 판단된다.

Radiation workers receive exposure during radiation works such as decontamination or cutting of metals and concrete in decommissioning nuclear power plants. To reduce occupational exposure, various radiation protection measures should be prepared by estimating the exposure dose in advance. RESRAD-RECYCLE, the computer code, is generally used for estimating occupational dose due to handling metals contaminated with radioactive materials. However, RESRAD-RECYCLE used the dose conversion factors (DCF) of EPA FGR No. 11 based on ICRP Publications 30 and 48 published in the 1980s for internal exposure estimation. This study compared the DCFs of RESRAD-RECYCLE with those of the relatively recently published ICRP Publications 119 and 141. In addition, the internal exposure dose was evaluated by changing the value of the DCFs of RESRAD-RECYCLE. As a result of the comparison, ICRP Publication 119 showed that the DCF values of most nuclides were significantly lowered. On the other hand, in the case of nuclides emitting gamma rays, there was generally no significant change in the value of DCFs. In addition, in the case of 65Zn and 94Nb, the DCF increased compared to the previous ICRP publications. The exposure dose of the decommissioning workers of Hanul Units 1 and 3 and Hanbit Unit 4 was also calculated in this study. The expected radioactivity concentration of the steam generator chamber of each unit was used as the source term. The concentration of metal dust in the air generated during cutting was calculated and applied to evaluate the internal exposure dose. As a result of the dose evaluation, there was a difference in exposure dose up to 0.2 mSv in the scrap cutter scenario of Hanbit Unit 4, which generated a lot of dust and had a high radioactivity concentration. On the other hand, in the case of the slag worker, there was no difference in the dose because the working time was very short, and the inhalation of metal dust was small, even if the latest DCF was applied.

3D modeling is a technology for representing real objects in a virtual 3D space or modeling and reproducing the physical environment. 2D drawings for viewing the existing building structure have limitations that make it difficult to understand the structure. By implementing this 3D modeling, specific visualization became possible. 3D technology is being applied in a wide range of preevaluation work required for nuclear decommissioning. In Slovakia, 3D modeling was applied to determine the optimal cutting strategy for the primary circuit before dismantling the VVER type Bohunice V1. In Japan, the Decommissioning Engineering Support System (DEXUS) program has been developed that incorporates VRDose, a decommissioning engineering support system based on 3D CAD models. Through this, the cutting length of the work object and the quantity of containers for packaging waste are calculated, exposure dose calculations in various dismantling scenarios, and cost estimation are performed. Korea also used 3D technology to evaluate the decommissioning waste volume for Kori Unit 1 and to evaluate the optimal scenario of the decommissioning process procedure for the research reactor Unit 1. 3D technology is currently being used in various pre-decommissioning evaluations for VVER, PWR, and research reactors. Overseas, a program that matches various decommissioning pre-evaluation tasks with cost estimation has also been developed. However, most 3D technologies are mainly used as a support system for dose evaluation and amount of decommissioning waste calculation. In this study, 3D modeling was performed on the PHWR structure, and physical and radiological information about the structure was provided. The information on the structure can present the unit cost for the work object by confirming the standard of the applied unit cost factor (UCF). The UCF presents the unit cost for repeated decommissioning operations. The decommissioning cost of the work object can be obtained by multiplying the UCF by the number of repetitions of the work. If the results of this study are combined with the process evaluation and waste quantity estimation performed in previous studies, it is judged that it will be helpful in developing an integrated NPP decommissioning program that requires preliminary evaluation of various tasks. In addition, it is judged that a clear cost estimation of the object to be evaluated will be possible by matching the 3D work object with the UCF.

For the decommissioning or continuous long-term power generation of nuclear power plants, it is necessary to transfer the spent nuclear fuel from the wet storage pool to the dry storage. Spent nuclear fuel should go through the drying process, which is the first step of dry storage. The most important part in the drying process is the removal of the residual water. The spent fuel might be stored in a dry storage system for a long time. The integrity of internal components and spent fuel cladding should be maintained during the storage period. If residual water is present, problems such as aging of metal materials, oxidation of cladding, and the hydride-reorientation could occur. The presence or absence of residual water after vacuum drying is evaluated by pressure. If there is residual water in the vacuum drying process, it evaporates easily at low pressure to form water vapor pressure and the internal pressure rises. In the recent EPRI High burn up demonstration test, the gas inside the canister that satisfied the dryness criteria was extracted and analyzed. It showed that the water content was higher than the expected value. We are conducting verification studies on the pressure evaluation method, which is an indirect evaluation method of vacuum drying. The vacuum drying test was performed on small specimens at Sandia National Laboratory, and quantitative residual water evaluation was also performed. The report did not mention a detailed method for the assessment of residual water. Based on the test results of SNL, direct residual water evaluation was performed using energy balance. If the dryness criteria were satisfied, the quantitative amount of residual water was also evaluated. As a result, almost the same result as the evaluation result of SNL was derived, and it was confirmed that most of the water was removed when the dryness criteria was satisfied.

Thermal analysis and safety assessment of spent fuel transport cask are mainly conducted using commercial Computational Fluid Dynamics (CFD) codes based on Finite Volume Method (FVM). The reliability and predictability of CFD codes have greatly been improved by the development in the computer systems, and are widely used to calculate heat flow in complex structures that cannot be analyzed theoretically. In the field of thermal analysis using the CFD code, it is important to clearly reflect the physical model of the transport cask, and a grid configuration suitable for the physical model is essential for accurate analysis. However, since there are no clear standard and guidelines for grid configuration and size, it is highly dependent on the user’s insight. Spatial discretization errors result from the use of finite-width grids and the approximation of the differential terms in the model equations by difference operators. Since the user usually cannot change the truncation error order of a given discretization scheme, spatial discretization errors can only be influenced by the provision of optimal grids. Therefore, it is necessary to quantify the spatial discretization errors caused by the grid. In the case of Orano TN’s NUHOMS® MP197 transport cask, considering four grids for two sets, the temperature uncertainty of the neutron shield, which has the lowest margin at the limit temperature among transport cask components, was quantified by applying 5-step procedure of the Grid Convergence Index (GCI) method for the uncertainty estimation presented in ASME V&V 20-2009. In the case of domestic spent nuclear fuel transport cask (KORAD21), neutron shield among the transport cask components has the lowest margin at the limited temperature. Accordingly, in this study, the temperature uncertainty of the neutron shield was quantified by applying GCI to three sets considering seven grids. As a result of the calculation, the uncertainty was less than ± 1°C, and the temperature of the neutron shield including the uncertainty was evaluated to be maintained below the limit temperature of 148°C.

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.

Concerns about North Korea’s 7th nuclear test have been rising recently, and it is a significant threat to the situation around the Korean Peninsula. Amidst these threats, the Korean government also shows a strong will for denuclearizing the Korean Peninsula, referring to the “Audacious Initiative.” For denuclearization negotiations with North Korea, it is essential first to understand North Korea’s nuclear capabilities. However, since access to information is complicated and contains many uncertainties, many studies have been conducted to estimate it. Among them, Von Hippel surveyed to estimate the total amount of uranium ore based on information on uranium mining, which is relatively widely known throughout North Korea’s nuclear fuel cycle, and the amounts of HEU and Pu suggested by many experts. KINAC has conducted a study on a methodology that can narrow the estimation range and improve reliability through the Bayesian Network based on Von Hippel’s research results. However, in this study, the probability distribution is assumed to be the simplest form of uniform distribution, and the estimation formula for the amount of Pu produced compared to the amount of uranium loaded in the core is used as it is, which is an error in Von Hippel’s study. Improvement is needed. This study proposes a more reliable BN model by supplementing this and attempts to estimate the amount of uranium ore that North Korea produces or possesses. Of course, the data used as the basic structure of the model is insufficient, and the estimation formula is straightforward, so it is somewhat unreliable to trust the estimate for uranium ore. However, it is expected to be a suitable methodology that can narrow the scope of North Korea’s nuclear material production estimate or compensate for the uncertainty of the nuclear material production estimation model being developed at KINAC.

PURPOSES : Roller-compacted concrete pavement (RCCP) is a superstiff-consistency concrete pavement that exhibits excellent strength development owing to a hydration reaction and interlocking aggregates owing to the roller compaction. A zero-slump concrete mixture is generally used. Hence, it is important to control the consistency of the RCCP mixture to prevent the deterioration of the construction quality (such as material separation during paving). The workability of the RCCP is characterized by its consistency and controlled by the Vebe time, whereas a conventional concrete pavement is controlled based on the slump test. The consistency of the RCCP changes over time after concrete mixing owing to delivery, construction time delays, etc. Thus, it is necessary to use the optimum Vebe time to achieve the best construction quality. Therefore, this study aims to develop a Vebe time prediction model for efficiently controlling the consistency of RCCPs according to random time variations. METHODS : A Vebe time prediction model was developed using a multiple linear regression analysis. A dataset of 131 samples was used to develop the model. The collected data consisted of variables with large potential effects on the consistency of the RCCP, such as the water-cement ratio (W/C), sand/aggregate ratio (S/a), water content (ω), water content per unit volume (W), cement (C), fine aggregate (S), coarse aggregate (G), water reducing admixtrue (PNS), air-entraining admixture (AE), delay time (T), air temperature (TEM), and humidity (HUM). In the multiple linear regression analysis, the mentioned parameters were used as the independent variables, and the Vebe time was the dependent variable. The Vebe time prediction models were evaluated by considering the adjusted R2 and p-values. The selection of the model was based on the largest R2 value and an acceptable p-value (p<0.05). RESULTS : The Vebe time prediction model achieved an adjusted R2 value of 64.14% with a significance level (p-value) of less than 0.05. This shows that the predictive model is adequately described for the dependent variable, and that the model is suitable for Vebe time predictions. Moreover, the significance level of the independent variables is less than 0.05, indicating significant effects on the Vebe time (i.e., the dependent variable). CONCLUSIONS : The Vebe time prediction model developed in this study can be used to estimate Vebe times with an R2 of 63.33% between the measured and predicted values. The proposed Vebe time prediction model is expected to be effectively utilized for the quality control of RCCP mixtures. Moreover, it is expected to contribute to achieving good RCCP construction quality.

스마트 항만시스템을 구축하기 위해서는 무엇보다 선박을 자동으로 계류시킬 수 있는 시스템이 요구된다. 항만의 자동계류시 스템의 허용 계류력을 산정하기 위해서는 선박의 특성을 고려하고, 해양환경적 외란으로 부터 발생된 외력을 정확히 계산해야 한다. 이러 한 환경적 외란의 크기를 정확히 추정하는 것은 자동계류장치 설계를 위해서 매우 중요한 요소이다. 본 연구에서는 항만 및 어항 설계기 준에 따라 한바다호에 대한 계류력을 추정하였다. 그 결과 한바다호에 작용하는 대부분의 외력은 바람으로부터 기인되는 것을 확인하였 다. 가장 극한 해양조건(B.F 6)에서 한바다호에 작용하는 종방향 힘은 18kN, 횡방향 힘은 248kN으로 나타났다.

In this study, we propose a flow velocity evaluation scheme based on pressure measurement in pressurized pipeline systems. Conservation of mass and momentum equations can be decomposed into mean and perturbation of pressure head and flowrate, which provide the pressure head and flowrate relationship between upstream and donwstream point in pressurized pipeline system. The inverse impedance formulations were derived to address measured pressure at downstream to evaluation of flow velocity or pressure at any point of system. The convolution of response function to pressure head in downstream valve provides the flow velocity response in any point of the simple pipeline system. Simulation comparison between traditional method of characteristics and the proposed method provide good agreements between two distinct approaches.