• Journal of Radiation Protection and Research
• /
• 제39권2호
• /
• pp.109-117
• /
• 2014

방사선비상계획은 원자력시설에 대한 사고해석을 통해 산출한 방사선원항과 기상자료에 근거한 선량평가 결과를 기초로 비상계획이 필요한 거리를 산출하고, 비상계획이 필요한 거리 내에 거주하고 있는 거주민에 대한 옥내대피, 소개, 갑상선방호 등의 보호조치 계획을 수립하는 방식으로 이루어진다. 본 연구에서는 연구용원자로 '하나로'에서 가상할 수 있는 최대사고 조건 하에서 부지내외 거주자에 대한 보호조치 전 후의 선량변화를 1년간 기상자료에 기초하여 확률론적으로 평가하고, 국제방사선방호위원회의 2007년 권고에서 제시한, 비상피폭상황에서 보호조치 이후 잔여선량으로 정의된 참조준위 개념을 사용하여, 최적의 보호조치 유형을 도출하였다. 하나로의 경우 최대 가상사고시 최적의 보호조치 유형은 반경 300 m 이내 거주자 소개, 반경 800 m 이내 거주자 옥내대피로 평가되었으며, 갑상선방호는 반경 600 m 이내 거주자에 국한하여 해당되는데 이 지역 거주자가 소개 또는 옥내대피시는 방사능방재요원 외에 필요가 없는 것으로 평가되었다.

• Journal of Radiation Protection and Research
• /
• 제41권4호
• /
• pp.368-372
• /
• 2016

Background: Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. Materials and Methods: It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. Results and Discussion: An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. Conclusion: The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

• Nuclear Engineering and Technology
• /
• 제36권1호
• /
• pp.24-35
• /
• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management

• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.937-946
• /
• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• 한국지반신소재학회논문집
• /
• 제22권3호
• /
• pp.27-35
• /
• 2023

굴착 공사 중 가시설의 안정성을 확보하는 것은 매우 중요하다. 설계 시 지반의 안정성을 수치해석을 통해 분석하고 있지만, 시공시에는 여건이 달라지기 때문에 계측으로 벽체 안정성을 분석하는 일은 필수불가결한 일이다. 공사현장에서의 계측비용은 매우 낮은 단가로 책정되어있으며 이를 통해 흙막이 벽체의 사고위험성은 예측되고 있다. 따라서 본 연구에서는 흙막이 벽체의 자동 혹은 무선 시스템 계측의 중요성을 가상의 사고사례 분석을 통해 공사기간 및 사고비용을 산정하고 이를 계측비용과 비교하여 무선 및 자동계측 업무의 중요성을 주장하였다. 굴착공사 중 중대형 파괴 시 사고처리 금액에 대하여 계측비용은 5% 미만으로 계측비용을 증가시켜 사고를 미연에 방지하는 것이 경제적일 수 있다.

• 대한안전경영과학회지
• /
• 제16권1호
• /
• pp.101-108
• /
• 2014

This study highlights the theme of human error of train drivers, conducting empirical analysis on the relationship between emergency stress, psychological fatigue, safety behavior, and accident. The hypothetical test results based on questionnaires received from 223 train drivers working at A subway firm indicate that emergency stress shows a significant positive effect on psychological fatigue, which in turn shows a significant negative influence on safety behavior. And safety behavior is shown having a significant negative relationship with accident. These results suggest the necessity of corporate-level approaches to depict the drastic causes of drivers' emergency stress, and to effectively manage this stress, as well as the necessity of making effort to enhance safety behavior, and to prevent or reduce accidents.

• Nuclear Engineering and Technology
• /
• 제41권7호
• /
• pp.929-944
• /
• 2009

The present study is concerned with the extension of the Effective Convectivity Model (ECM) to the phase-change problem to simulate the dynamics of the melt pool formation in a Light Water Reactor (LWR) lower plenum during hypothetical severe accident progression. The ECM uses heat transfer characteristic velocities to describe turbulent natural convection of a melt pool. The simple approach of the ECM method allows implementing different models of the characteristic velocity in a mushy zone for non-eutectic mixtures. The Phase-change ECM (PECM) was examined using three models of the characteristic velocities in a mushy zone and its performance was compared. The PECM was validated using a dual-tier approach, namely validations against existing experimental data (the SIMECO experiment) and validations against results obtained from Computational Fluid Dynamics (CFD) simulations. The results predicted by the PECM implementing the linear dependency of mushy-zone characteristic velocity on fluid fraction are well agreed with the experimental correlation and CFD simulation results. The PECM was applied to simulation of melt pool formation heat transfer in a Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lower plenum. The study suggests that the PECM is an adequate and effective tool to compute the dynamics of core melt pool formation.

• 방사성폐기물학회지
• /
• 제19권3호
• /
• pp.307-322
• /
• 2021

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

• 방사성폐기물학회지
• /
• 제4권2호
• /
• pp.139-152
• /
• 2006

사용후핵연료 건식저장용기는 낙하사고조건에서 캐니스터의 건전성이 입증되어야 한다. 낙하사고조건은 캐니스터를 건식저장용기에 장입하기 위하여 저장용기의 상부에서 크레인으로 취급하는 도중에 캐니스터가 저장용기 내부의 받침대로 자유 낙하하는 조건이다. 저장용기 내부의 받침대는 이러한 조건에서 캐니스터의 구조적 건전성을 유지하도록 완충효과가 좋아야 한다. 본 연구에서는 다양한 저장용기 내부 받침대 에 대한 3차원 유한요소해석을 통하여 낙하사고조건에서 캐니스터의 구조적 건전성을 향상시킬 수 있는 구조를 결정하였다. 저장용기 내부 받침대는 탄소강으로 만들어진 원통 쉘의 내부에 콘크리트를 장입한 구조와 받침대 높이의 변화 없이 콘크리트 높이의 1/4정도에 탄소강과 폴리우레탄폼을 이용한 구조물을 사용하여 완충효과를 보완하고자 수정된 구조를 고려하였다. 완충체의 형상 및 구조를 결정하기 위하여 십자형상이나 원형의 탄소강 구조물을 받침대 상부에 위치하여 그 영향을 알아보았다. 이때 탄소강 구조물의 두께를 24 mm, 12 mm, 6mm로 변화를 주었다. 또한, 탄소강 구조물 사이에 충진하는 폴리우레탄폼의 밀도에 대한 영향을 알아보았다.

• Nuclear Engineering and Technology
• /
• 제44권6호
• /
• pp.597-610
• /
• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.