• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.657-662
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• 1996

Venting capability to prevent excess pressurization caused by loss of Residual Heat Removal System (RHRS) during mid-loop operation hat been evaluated analytically and the peak Reactor Coolant System (RCS) pressure was compared with the results of the MIDLOOP computer code. Even though analytical method if relatively simple, the results are in a good agreement with those of the computer code. For both methods, the peak pressures have not, exceeded the nozzle dam design pressure, if the vent paths such as pressurizer safety valves or a pressurizer manway are available in a closed RCS configuration with the nozzle dam installed.

• 한국안전학회지
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• 제21권4호
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• pp.49-54
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• 2006

The evaluation of elastic-plastic fracture characteristic was investigated in ferrite steel SA 516- Gr70 used for reactor coolant piping elbow and support skirt of pressure vessels. This paper describes the effect of temperature on J-R curve characteristic of this material. The elastic-plastic fracture mechanics parameter J is obtained with unloading compliance method. The test method were analyzed according to ASTM E 813-89 and E 1152-89. Unloading compliance $J_{IC}$ tests were performed on 1 CT specimens at varied temperatures from $25^{\circ}C$ to about $400^{\circ}C$ using a high temperature extensometer. At all temperature, valid $J_{IC}$ measurements could be made and $J_{IC}$ decreased with increasing temperature. SEM fractography schematically illustrates microvoid initiation, growth and coalescence at the tip of a preexisting crack.

• 방사성폐기물학회지
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• 제16권3호
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• pp.389-396
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• 2018

국내 가동원전 중 2-루프 가압경수로인 고리1호기는 약 40년 운전한 후, 2017년 6월 18일 영구정지되었다. 영구정지된 고리 1호기는 주요 해체작업을 수행하기전에 계통내 선량률을 저감시켜 작업자피폭을 최소화하기 위한 계통제염을 수행할 예정이다. 일반적으로, 계통제염 범위는 원자로압력용기, 가압기, 증기발생기, 화학 및 체적제어계통, 잔열제거계통 및 원자로 냉각재계통 주요배관을 포함한다. 이러한 계통 및 기기 등을 효율적으로 제염하기 위해서는 제염과정에서 원자로냉각재계통내 유동특성을 평가할 필요가 있다. 계통제염을 위해 순환유량을 제공하는 방법은 다양하나, 본 논문에서는 잔열제거펌프 운전에 따른 고리1호기 원자로냉각재계통내 유동특성을 평가하였다. 잔열제거펌프를 이용한 계통제염은 원자로냉각재 내 유량의 불균형을 초래하여 계통내 기기 및 배관 등에 불순물을 침적시켜 제염이 효율적이지 않다는 것으로 평가되었다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.813-820
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• 2001

The Safety depressurization System(SDS) of KNGR prevents RCS from overpressurization by discharging high pressure and temperature coolant through the I-sparger into the IRWST during an accident. If IRWST water temperature rise locally, around the sparger, beyond $200_{\circ}$2000 F by the discharged coolant, unstable steam condensation can cause large pressure load on the IRWST wall. To investigate whether this condition can be avoided for the design basis event IOPOSRV(Inadvertent Opening of one Pilot Operated Safety Relief Valve), the flow and temperature distribution of water in the IRWST is calculated by using CFX 4.3 computational fluid dynamic code. According to the results, since pool water temperature does not exceeds temperature limit within 50 seconds after the opening of one POSRV, it can be assured that the integrity of IRWST wall is maintained.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3567-3579
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• 2022

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

• 한국산학기술학회논문지
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• 제5권3호
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• pp.227-231
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• 2004

It is the common features of the integral reactors that the main components of the RCS are installed within the reactor vessel, and so there are no any flow pipes connecting the steam generator or the pump whose type is the axial flow. Due to no any flow pipes, it is impossible to measure the differential pressure at the RCS of the integral reactors, and it also makes impossible measure the flow-rate of the reactor coolant. As a alternative method, the method by the measurement of the pump power of the axial flow pump has been introduced in this study. Up to now, we did not found out a precedent which the pump power is used for the flow-rate calculation at normal operation of the commercial nuclear power plants. The objective of the study is to embody the flow-rate calculation method by the measurement of the pump power in an integral reactor. As a result of the study, we could theoretically reason that the capacity-head curve and capacity-shaft power curve around the rated capacity with the high specific-speeded axial flow pumps have each diagonally steep incline but show the similar shape. Also, we could confirm the above theoretical reasoning from the measured result of the pump motor inputs. So, it has been concluded that it is possible to calculate the flow-rate by the measurement of the pump motor inputs.

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

• Journal of Radiation Protection and Research
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• 제43권4호
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• pp.154-159
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• 2018

Background: To prevent small leakage accidents, a real-time and direct detection system for small leaks with a detection limit below that of existing systems, e.g. $0.5gpm{\cdot}hr^{-1}$, is required. In this study, a small-size beta detector, which can be installed inside the reactor containment (CT) building and detect small leaks directly, was suggested and its feasibility was evaluated using MCNPX simulation. Materials and Methods: A target nuclide was selected through analysis of radiation from radionuclides in the reactor coolant system (RCS) and the spectrum was obtained via a silicon detector simulated in MCNPX. A window was designed to reduce the background signal caused by other nuclides. The sensitivity of the detector was also estimated, and its shielding designed for installation inside the reactor CT. Results and Discussion: The beta and gamma spectrum of the silicon detector showed a negligible gamma signal but it also contained an undesired peak at 0.22 MeV due to other nuclides, not the $^{16}N$ target nuclide. Window to remove the peak was derived as 0.4 mm for beryllium. The sensitivity of silicon beta detector with a beryllium window of 1.7 mm thickness was derived as $5.172{\times}10^{-6}{\mu}Ci{\cdot}cc^{-1}$. In addition, the specification of the shielding was evaluated through simulations, and the results showed that the integrity of the silicon detector can be maintained with lead shielding of 3 cm (<15 kg). This is a very small amount compared to the specifications of the lead shielding (600 kg) required for installation of $^{16}N$ gamma detector in inside reactor CT, it was determined that beta detector would have a distinct advantage in terms of miniaturization. Conclusion: The feasibility of the beta detector was evaluated for installation inside the reactor CT to detect small leaks below $0.5gpm{\cdot}hr^{-1}$. In future, the design will be optimized on specific data.

• 에너지공학
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• 제26권3호
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• pp.71-77
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• 2017

피동고압충수용 혼합형 안전주입탱크 (Hybrid SIT)의 압력평형시간은 냉각수 주입시기를 결정하는 주요인자이다. 한국원자력연구원 (KAERI)에서는 Hybrid SIT에서의 내부 열수력적 거동을 고찰하기 위해 개별효과시험 장치를 구축하였으며, 다양한 운전조건에서의 압력평형시간에 대한 민감도 시험을 수행하였다. 개별효과시험을 통해 압력평형시간을 결정하는 주요인자들을 도출하였으며, 그 중 증기의 벽면응축 및 냉각재와의 직접접촉응축이 압력평형시간을 결정하는 주요 현상임을 파악하였다. 본 연구에서는 개별효과 시험결과들을 이용하여 각각의 응축현상들이 압력평형에 미치는 영향을 정량적으로 분석하고 혼합형 SIT의 압력평형시간을 예측하기 위한 방법론을 제시하였다.