• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1651-1664
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• 2023

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2349-2355
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• 2023

In the TRITIUM project, an on-site monitoring system is being developed to measure tritium (³H) levels in water near nuclear power plants. The quite low-energy betas emitted by ³H have a very short average path in water (5 ㎛ as shown by simulations for 18 keV electrons). This path would be further reduced by impurities present in the water, resulting in a significant reduction of the detection efficiency. Therefore, one of the essential requirements of the project is the elimination of these impurities through a filtration process and the removal of salts in solution. This paper describes a water treatment system developed for the project that meets the following requirements: the water produced should be of near-pure water quality according to ISO 3696 grade 3 standard (conductivity < 10 µS/cm); the system should operate autonomously and be remotely monitored.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• 핵의학기술
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• 제17권1호
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• pp.53-61
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• 2013

방사성 요오드 갑상선 섭취율은 거대갑상선 환자의 경우 그 체적에 의한 유효 갑상선 깊이가 깊어짐으로 인한 기하학적 변동이 있는 것이 사실이다. 본 연구는 방사성 요오드갑상선 섭취율에 있어 검출기와 선원 간 거리와 유효 갑상선 깊이에 따른 기하학적 요인의 영향을 고찰하고자 하였다. $^{131}I$ 370 kBq 선원을 검출기로부터 거리를 20 cm부터 30cm까지 1 cm 간격으로 변화시키며 Captus 3000 thyroid uptake system(Capintec, NJ, USA)으로 측정하였다. 유효갑상선 깊이를 재현하기 위해 목 팬텀을 이용하여 팬텀 내 선원의 깊이를 1 cm, 2 cm, 5 cm으로 변화시키며 같은 방법으로 측정하였다. 실험 결과, 곡선추정 회귀분석 결과 모든 실험군이 거듭제곱곡선에 높은 상관관계를 보이는 것으로 나타났다($$R2{\geq_-}0.915$$). 그러므로 검출기-선원 간 거리가 20 cm보다 30 cm에서 오차가 크게 감소됨을 예상할 수 있다. 모든 실험군에서 팬텀을 쓰지 않았을 때와 유효 갑상선깊이가 1 cm이 적용되었을 때의 계수율이 서로 유의할 만한 차이가 있는 것으로 나타났다(p<0.01). 선형회귀분석 결과 깊이에 따른 계수율의 변화는 모두 감소되는 것으로 나타났으나,$284.3keV{\pm}10%$ 영역에서 깊이에 따른 계수율의 변화는 증가되는 것으로 나타났다. 이 회귀식을 통해 환자의 예상 갑상선 섭취율을 산출해 보았을 때, $364.4keV{\pm}10%$에서 1 cm 당 -6.42%, $364.4keV{\pm}20%$의 영역에 서 -5.09%의 더 낮은 오차를 보였다. 또한 거리에 따른 계수율의 변동계수는 모든 실험군에서 선형으로 증가되는 것으로 나타났다. 그 중 $364.4keV{\pm}20%$, $364.4keV{\pm}10%$ 영역은 비교적 낮은 변동계수와 증가폭을 보였다. 곧, 유효 갑상선 깊이에 따른 오차를 줄이기 위해서는 $364.4keV{\pm}20%$의 영역의 사용이 더 적절할 것으로 보인다. 그러므로 갑상선 깊이에 따른 오차는 갑상선 깊이에 따른 보정계수 적용,$364.4keV{\pm}20%$ 에너지 영역 설정, 디텍터와 선원과의 거리를 연장하였을 때 감소시킬 수 있다고 생각된다.

• 핵의학기술
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• 제14권1호
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• pp.40-45
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• 2010

Purpose: Recently, the performance of the X-ray tube was very much improved by the power generation of the technology. However, the overload of equipment is occurred by the increment of the equipment operating time according to the increment of the examination number of cases. The X-ray dose can change by heat occurrence of the X-ray tube due to this. Moreover, the change of the bone mineral density value is possible to occur. Therefore, We tries to whether the change of the bone mineral density value of each equipment according to the difference of the examination number of cases and operating time occur or not. Materials and Methods: The BMD value was measured by the Aluminum Spine Phantom and the European Spine Phantom in each equipment, in order to find out about the difference of the time general classification bone mineral density value by using the Dual energy X-ray absorptiometry. And after scanning each phantom by using X-ray dose meter (Unfors Mult-O-Meter), a dose was measured by the same condition. As to, an average and standard deviation were found and the change of each equipment much BMD value was compared and it evaluated. Results: $Mean{\pm}SD$ of each equipment by using the Aluminum Spine Phantom, A equipment was $1.174{\pm}0.002$, $1.171{\pm}0.005$, $1.173{\pm}0.005$, B equipment was $1.186{\pm}0.003$, $1.187{\pm}0.003$, $1.185{\pm}0.003$, C equipment was $1.180{\pm}0.003$, $1.182{\pm}0.004$, $1.183{\pm}0.002$, D equipment was $1.188{\pm}0.004$, $1.185{\pm}0.003$, $1.185{\pm}0.004$. By using the European Spine Phantom, A equipment was $1.143{\pm}0.006$, $1.153{\pm}0.009$, $1.161{\pm}0.003$, B equipment was $1.134{\pm}0.004$, $1.13{\pm}0.008$, $1.127{\pm}0.015$, C equipment was $1.143{\pm}0.006$, $1.134{\pm}0.01$, $1.133{\pm}0.006$, D equipment was $1.14{\pm}0.001$, $1.122{\pm}0.002$, $1.131{\pm}0.008$, altogether included in the normal range. Conclusion: There was no significant change of the BMD value of using a phantom by time zones. Therefore, if the quality control is made to use the extent management method of the equipment for beginning in the present application, the reliability of the BMD equipment will be able to be enhanced.

• 터널과지하공간
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• 제19권6호
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• pp.490-497
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• 2009

최근 발생한 5개의 중규모 지진으로부터 관측된 지반진동 파형을 이용하여 응답스펙트럼을 분석하고 결과를 국내 원자력 관련 구조물의 내진설계 기준과 국내 일반 구조물 및 건축물 내진설계기준과 각각 비교하였다. 연구에 이용된 지반진동 개수는 수평성분 및 수직성분 각각 74개 및 89개이며 주파수별 지반응답을 구하고 최대 지반 가속도 값를 이용하여 정규화 분석을 수행하였다. 본 연구결과를 국내 원자력시설물의 내진기준으로 이용되고 있는 Reg. Guide 1.60과 비교한 결과 특히 약 1 Hz 이상의 전체 고주파수 영역에서 수평 성분 스펙트럼 이 Reg. Guide 1.60 보다 높은 값을 보여 주었다. 수직성분은 약 7~8 Hz 부근에서 약간 초과하였다. 또한 국내 일반 구조물 및 건축물 내진설계기준인 표준 설계응답스펙트럼을 3개 지반조건에 적용한 결과를 분석 자료와 동시에 비교한 결과 특히 약 2초(0.5 Hz) 이하의 단주기 영역의 전체 대역(SE 지반조건)에서 수평 성분 자료처리 결과가 기준을 크게 초과하는 현상을 보여 주었다. 수직성분은 전체 주기 영역에서 SD 지반조건의 기준과 유사한 특징을 보여 주었다. 물론 이러한 현상은 국내 지각의 주파수별 감쇠 및 부지 직하부의 감쇠 특성 등과 복합적으로 관련되어 발생한 현상으로 판단된다. 향후 국내 지진활동 실정에 적합한 내진설계 기준 마련을 위해 관측자료의 질적 향상 및 양적인 축적 등을 통하여 특히 수평 성분의 약 1 Hz 이상의 고주파수 대역에서 응답스펙트럼 기준의 보수성을 심각하게 고려할 필요가 있다.

• Journal of Radiation Protection and Research
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• 제44권4호
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• pp.128-148
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• 2019

Massive environmental monitoring has been conducted continuously since the Fukushima Daiichi Nuclear Power accident in March of 2011 by different monitoring methods that have different features together with migration studies of radiocesium in diverse environments. These results have clarified the characteristics of radiological environments and their temporal change around the Fukushima site. At three months after the accident, multiple radionuclides including radiostrontium and plutonium were detected in many locations; and it was confirmed that radiocesium was most important from the viewpoint of long-term exposure. Radiation levels around the Fukushima site have decreased greatly over time. The decreasing trend was found to change variously according to local conditions. The air dose rates in environments related to human living have decreased faster than expected from radioactive decay by a factor of 2-3 on average; those in pure forest have decreased more closely to physical decay. The main causes of air dose rate reduction were judged to be radioactive decay, movement of radiocesium in vertical and horizontal directions, and decontamination. Land-use categories and human activities have significantly affected the reduction tendency. Difference in the air dose rate reduction trends can be explained qualitatively according to the knowledge obtained in radiocesium migration studies; whereas, the quantitative explanation for individual sites is an important future challenge. The ecological half-lives of air dose rates have been evaluated by several researchers, and a short-term half-life within 1 year was commonly observed in the studies. An empirical model for predicting air dose rate distribution was developed based on statistical analysis of an extensive car-borne survey dataset, which enabled the prediction with confidence intervals. Different types of contamination maps were integrated to better quantify the spatial data. The obtained data were used for extended studies such as for identifying the main reactor that caused the contamination of arbitrary regions and developing standard procedures for environmental measurement and sampling. Annual external exposure doses for residents who intended to return to their homes were estimated as within a few millisieverts. Different forms of environmental data and knowledge have been provided for wide spectrum of people. Diverse aspects of lessons learned from the Fukushima accident, including practical ones, must be passed on to future generations.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 추계학술대회논문집
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• pp.84-91
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• 2003

Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed. From these results, the current design of the APR1400 steam generator are examined.

• 방사선산업학회지
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• 제17권3호
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• pp.275-282
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• 2023

Workers in nuclear power plants are likely to be exposed to radiation from various geometrical sources. In order to evaluate the exposure level, the point-kernel method can be utilized. In order to perform a dose assessment based on this method, the radiation source should be divided into point sources, and the number of divisions should be set by the evaluator. However, for the general public, there may be difficulties in selecting the appropriate number of divisions and performing an evaluation. Therefore, the purpose of this study is to develop an algorithm for dose assessment for arbitrary shaped sources based on the point-kernel method. For this purpose, the point-kernel method was analyzed and the main factors for the dose assessment were selected. Subsequently, based on the analyzed methodology, a dose assessment algorithm for arbitrary shaped sources was developed. Lastly, the developed algorithm was verified using Microshield. The dose assessment procedure of the developed algorithm consisted of 1) boundary space setting step, 2) source grid division step, 3) the set of point sources generation step, and 4) dose assessment step. In the boundary space setting step, the boundaries of the space occupied by the sources are set. In the grid division step, the boundary space is divided into several grids. In the set of point sources generation step, the coordinates of the point sources are set by considering the proportion of sources occupying each grid. Finally, in the dose assessment step, the results of the dose assessments for each point source are summed up to derive the dose rate. In order to verify the developed algorithm, the exposure scenario was established based on the standard exposure scenario presented by the American National Standards Institute. The results of the evaluation with the developed algorithm and Microshield were compare. The results of the evaluation with the developed algorithm showed a range of 1.99×10^-1~9.74×10^-1 μSv hr^-1, depending on the distance and the error between the results of the developed algorithm and Microshield was about 0.48~6.93%. The error was attributed to the difference in the number of point sources and point source distribution between the developed algorithm and the Microshield. The results of this study can be utilized for external exposure radiation dose assessments based on the point-kernel method.

• 한국건설관리학회논문집
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• 제8권1호
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• pp.66-77
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• 2007

2001년 4월 정부의 전력산업 구조개편으로 한국전력공사는 한국수력 원자력과 5개 화력발전사로 분할되었다. 이후 공기업 민영화를 위한 각 발전회사들 간의 경쟁체제에 따라 적극적인 수익창출을 위하여 다양한 발주방식 도입, 적용하고 있다. 하지만, 화력발전소 건설사업은 발주자의 사업목표나 프로젝트의 특성을 제대로 반영하지 못한 채 발주방식을 선정함으로써 프로젝트 참여자들을 만족시키지 못하고 있는 실정이다. 본 연구는 화력발전소 건설사업에서의 발주방식을 선정하기 위해서 고려해야 할 영향요인들을 조사하고 설문 및 면담조사를 통하여 적정한 발주방식의 선정기준을 개발함으로써 발주자의 사업목표에 맞는 발주방식을 선정할 수 있는 모형을 구축하는 것을 목적으로 한다. 향후 본 연구에서 제안한 발주방식 선정모형을 활용하면 주관적이고 경험적인 판단에 의존하여 발주방식을 결정했던 기존의 관행을 벗어나 발주자의 사업목표, 발주자의 특성 및 요구조건, 건설사업의 특성 등을 효과적으로 반영하여 화력발전소 건설사업의 적정한 발주방식을 선정할 수 있을 것으로 기대된다.