For the release of the nuclear power plant site after the decommissioning, a reliable exposure dose assessment considering the environmental impact of residual radionuclides is essentially required. In this study, the Derived Concentration Guideline Level (DCGL) for the hypothetically contaminated surface soil at the Wolsong nuclear power plant (NPP) unit 1 site was preliminarily calculated by using the RESRAD-OFFSITE computational code and compared with the other case studies. Moreover, radiation exposure dose for local residents and relevant exposure pathways were quantitatively analyzed based on the calculation model established through this work. For the target site modeling, the source term was determined by referring to the previous case studies regarding the nuclear power plant decommissioning, quantification analysis data of pressure tubes of Wolsong NPP unit 1, and radionuclide data estimated by using the MCNP/ORIGEN-2 code. In total, 14 different radioisotopes such as Ag-108m, C-14, Co-60, Cs-134/137, Fe-55, H-3, Nb-93m/94, Ni-63, Sb-125, Sn-121m, Sr-90, and Zr-93 were considered as target radionuclides. In addition, the geological structure model of the Wolsong NPP site was established based on the final safety analysis report of Wolsong NPP unit 1. The distribution coefficients (Kd) were taken from the JAEA-SDB to estimate the migration/retardation behavior of various radionuclides under the groundwater condition of the Wolsong NPP site. In the present work, the DCGL values were calculated according to the site release criterion of 0.1 mSv/yr, which indicates the radiation protection standard for the site release. Moreover, the exposure pathway and sensitivity analyses were conducted to assess the sensitive input parameters remarkably influencing the calculation result. For the evaluation of exposure dose for local residents, a site layout centered around Wolsong NPP unit 4, located in the closest proximity to the residents’ habitation area, was alternatively established and all potential exposure pathways were considered as a comprehensive resident farmer scenario. The results obtained from this study are expected to serve as a preliminary case study for the DCGL values regarding the surface soil at the Wolsong NPP unit 1 site and for evaluating the radiation exposure dose to local residents resulting from the residual radioactivity at the site after the decommissioning.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

When a radiation detector is applied to the measurement of the radioactivity of high-level of radioactive materials or the rapid response to the nuclear accident, several collimators with the different inner radii should be prepared according to the level of dose rate. This makes the in-situ measurement impractical, because of the heavy weight of the collimator. In this study, an IRIS collimator was developed so as to have a function of controlling the inner radius, with the same method used in optical camera, to vary the attenuation ratio of radiation. The shutter was made to have the double tungsten layers with different phase angles to prevent the radiation from penetrating owing to the mechanical tolerance. The performance evaluation through the MCNP code was conducted by calculating the attenuation ratio according to the inner radius of the collimator. The attenuation ratio was marked on the outer scale ring of the collimator. It is expected that when a radiation detector with the IRIS collimator is used for the in-situ measurement, it can change the attenuation ratio of the incident photon to the detector without replacing the collimator.

이 연구에서는 안전관리체제(ISM Code) 실행지원 모듈개발중 하나인 ISM Code 운영평가 모듈 개발을 위한 기초적 연구를 수행하였다. 특히 ISM Code 운영평가 모듈 개발을 중심으로 AHP의 운영평가 분석과정과 설문지 설계, 운영평가 모형 기준의 위계 구조를 정립하였다는 것에 대해 그 의미가 있겠다. 또한 경영과학 분야의 주요 의사결정 하나인 AHP 기법을 운영평가 모듈 개발에 적용을 시도했다는 것에 대해서도 한층 가치가 있겠다.

본 연구에서는 지반상태가 교량의 지진응답에 미치는 영향을 분석하기 위해 다양한 지반모델 상의 교량모델에 대한 지진해석을 수행하고 그 결과를 비교.분석하였다 이를 위해서 기존의 교량시방서에서 분류된 네가지의 서로다른 지반종류를 대상으로 보다 세분된 지반모델 상에 위치하는 대표적인 교량에 대한 지진해석을 수행함으러써 첫째로는 각 지반 종류 상호간의 지진응답 차이의 정도를 시방서 값과 비교.분석하였으며 둘째로는 동일 지반으로 분류되는 상이한 두 지반간의 응답차이의 정도를 확인하였다 해석을 위해서는 시방서에서 제시된 두가지 방법 즉 단일모드 스펙트럼해석법과 다중모드 스펙트럼해석법을 사용하였으며 이들 결과를 별도로 작성된 인공시간이력을 입력으로 하는 시간이력해석법을 사용한 결과와 비교.분석하였다 시간이력해석법에서는 진동수에 무관한 지반임피던스함수를 이용하는 시간영역해석법을 사용하였다 해석결과 시방서에서 제시한 단일모드 및 다중모드 스펙트럼해석방법은 일반적으로 안전측의 지반-구조물 상호작용 해석결과를 주는 것으로 확인되었다 그러나 유연성이 큰 지반 상의 구조물에대한 지진해석을 위한 해석모델 작성시에는 지반의 유연성이 반드시 고려되어야 하며 특히 낙교방지를 위한 변위는 지반의 유연성을 고려한 정밀해석법에 의한 계산이 필수적으로 요구됨을 확인하였다.

현행 내진설계 규준에서 사용하고 있는 반응수정계수는 설계지진하중과 유사한 지진발생시 구조물이 비선형 거동을 하도록 탄성응답에서 요구되는 밑면전단력 값을 낮추는 계수라 할 수 있다. 따라서 반응수정계수는 하중저감계수(force reduction factor)라고 할 수 있으며, 이러한 값들은 경험적으로 결정된 것이어서 예상지진에 대하여 구조설계자가 설계한 건물이 어느정도의 비선형 거동을 할지는 예측하기가 힘들다. 본 연구에서는 목표가 되는 연성비(target ductility ratio)에 따라 요구되는 밑면전단력의 값을 구하고 이를 규준에서 요구하는 값과 비교할 것이다. 만약 요구되는 값이 규준 값 보다 크다면 이는 구조물이 가지는 부가강도(overstrength)나 잉여력(redundancy)이 담당해야 한다. 모멘트연성골조 건물을 설계한 후 이를 push-over 해석에 의하여 부가강도를 찾아 보아 요구강도와 비교할 것이다.

본 연구는 표준 감마선 조사장치의 유효빔 크기를 실측과 시뮬레이션의 결과를 비교하여 유효빔 영역의 결정에 유용한 수단을 제공하고자 하였다 시뮬레이션과 전리함을 이용한 실측의 결과는 공기커마율의 경우는 상대오차 4.5~7.3% 범위에 분포하였다. 유효빔 영역의 크기는 시뮬레이션의 경우 수평 방향 27cm, 수직 방향 21.6cm로 구현되었고, 필름을 이용한 실측결과는 수평 방향 26.5cm, 수직 방향 21.9cm로 유사한 결과가 도출되었다. 수평방향의 상대오차는 1.85%, 수직 방향은 1.38% 이며 유효빔 영역도 감마선장을 중심으로 유사하게 분포하였다. 감마선 조사장치에 있어서 시뮬레이션의 유효성이 충분함을 확인하였다. 특히 공기커마율보다 유효빔 크기의 상대오차가 적은 것은 빔의 크기가 표준선원의 용량보다는 기하학적 요인으로 결정되기 때문인 것으로 판단된다. 향후 시뮬레이션을 이용한 광자 에너지 분포도의 신뢰성을 높이기 위한 연구가 필요 할 것이다.

This study examined the safety and reliability of code equations for the bond strength of FRP bars with concrete using test data compiled from the different sources. ACI equations have a better agreement with test results, whereas high overestimation is observed in JSCE equations.

Recently, Korea Bridge Design Code(limit state design method) was introduced in 2012.However, Korea Infrastructure’s Safety Inspection Guideline are still following the bridge safety assessment. The assessment methods are classified according to material and bridge superstructure. Steel box girder bridge safety assessment was carried out by the A.S.D. In this study, introduce of steel box girder bridge safety assessment in accordance with the standards change.

Recently, Korea Bridge Design Code(KBDC) adapts to the Limit State Design method(L.S.D) in 2012 according to the liberalization of construction market and the flow of international standardization of the design criteria. However, Korea Infrastructure’s Safety Inspection Guideline(KISIG) are still following the bridge safety evaluation applying the A.S.D and U.S.D methods. In this study, introduction of PSC bridge safety evaluation with L.S.D considered the traditional safety evaluation compared to the ASD and USD methods.

The Recent earthquake damage of surrounding countries is increasing, occurrence of frequent earthquakes in the country, thus, the interest about the seismic performance of the structure is increasing. Korea, was established in 2008 earthquake measures method, KBC2009 was also amended by design standards. KBC2009 based seismic design and seismic strengthening has become mandatory, Design of the building before the current seismic design standards must be satisfied in KBC2009. Therefore, ATC-40, FEMA-273 and seismic performance evaluation of Korea Infrastructure Safety Corporation will conduct the seismic performance evaluation at the Previous design code building.