• Nuclear Engineering and Technology
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• 제11권3호
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• pp.195-202
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• 1979

In this paper, a cause-consequence chart is applied to evaluate the probability that the containment spray system in a nuclear power plant may not be woring properly, given a demand for spryaing at loss of coolant accident (LOCA). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of this system, in case the test policy of the system is taken into account, i.e., average probability that the containment spray cannot be established, and average probability that the containment spray is established : spray stops before the required operating time is over.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.328-333
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• 1996

Boric acid concentrations of the refueling water storage tank (RWST) and the accumulators for Ulchin 1&2 (UCN 1&2) are increased to meet the post loss of coolant accident (post-LOCA) shutdown requirement for the extended fuel cycles from 12 months to 18 months. To maintain long term cooling (LTC) capability following a LOCA, the switchover tine is examined using BORON code to prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results show that, at 8 hours after the initiation of LOCA. the emergency core noting system (ECCS) should be manually realigned to the simultaneous recirculation mode from the cold leg recirculation mode.

• Nuclear Engineering and Technology
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• 제30권1호
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• pp.47-57
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• 1998

Boric acid concentrations of the refueling water storage tank and the accumulators for Westinghouse 3-loop type plants are increased to meet the post loss of coolant accident shutdown requirement for the extended fuel cycles from 12 months to 18 months. To maintain long term cooling capability following a LOCA, the switchover time is examined using BORON code to prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results show that hot leg recirculation switchover times are shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3&4 and 8 hours from 18 hours for Ulchin 1&2, respectively. The How path in the mode J for Kori 3&4 is recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1&2.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.475-479
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• 1995

During the hypothetical loss of coolant accident in the nuclear power plant the emergency core cooling water could not penetrate to the reactor core when the steam flow rate from the reactor core exceeds CCFL (Countercurrent flow limitation). The CCFL generated by earlier investigators are developed under the steady gas flow. However the flow instability in the reactor loop could generate oscillatory steam flow, hence their applicability under oscillating flow should be investigated. In this work, an experimental investigation of countercurrent flow in the vertical flow channel has been conducted under oscillatory gas flow. Pulsation of gas under oscillatory flow disturbs the flow pattern significantly and prevents flooding (CCFL) when its minimum value is less than the threshold gas flow rate value.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2006년도 춘계학술대회논문집
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• pp.246-249
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• 2006

Finite element vibration analysis of the trial 5x5 partial fuel assembly in the still water was performed using the substructure method. ANSYS software was used as a finite element modeling and modal analysis tool. The calculated natural frequencies of the partial fuel assembly were more consistent with the experimental results for the identical test model compared to the much larger solid model. This modeling technique can be utilized for the fuel assembly dynamic behavior analysis under normal operation, seismic and loss-of-coolant-accident analysis.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1994년도 추계학술발표회 초록집
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• pp.141-145
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• 1994

A computer code for solving the 3-D time-dependent multigroup neutron diffusion equation by a coupled reactor kinetics method recently developed has been developed and for evaluating its applicability in CANDU transient analysis applied to a 3-D kinetics benchmark problem which reveals non-uniform loss of coolant accident followed by an asymmetric insertion of shutdown devices. The performance of the method and code has been compared with the CANDU design code, CERBERUS, employing a finite difference improved quasistatic method.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2295-2303
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• 2021

To investigate wall-thinning impact on axial load resistance of Zircaloy-4 cladding rods after a LOCA transient, axial tensile samples have been machined on as-received tubes with reduced thicknesses between 370 and 580 ㎛. After high temperature oxidation under steam at 1200 ℃ with measured ECR ranging from 10 to 18% and water quenching, machined samples were axially loaded until fracture. These tests were modeled using a fracture mechanics approach developed in a previous study. Fracture stresses are rather well predicted. However, the slightly lower fracture stress observed for wall-thinned samples is not anticipated by this modeling approach. The results from this study confirm that characterizing the axial load resistance using semi-integral tests including the creep and burst phases was the best option to obtain accurate axial strengths describing accurately the influence of wall-thinning at burst region.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.597-609
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• 2020

Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr-1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, Cr_xN_y layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360℃ in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400℃ related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.

• 방사성폐기물학회지
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• 제9권4호
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• pp.231-236
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• 2011

지난 2011년 3월의 후쿠시마 원전 사고시 원자로 건물에서의 연쇄적인 수소폭발이 발생하였을 때 관계자들은 제1원전 4호기의 폭발에 더욱 놀랐었는데 이는 그 당시 4호기는 정기보수를 위하여 원자로내 모든 핵연료를 저장조에 보관중이었기 때문이다. 저장조내 냉각수 유실로 노심에서 옮겨진 핵연료가 공기 중에 노출되어 수소가 발생하고 임계가 도달하였다면 더욱 심각할 수도 있기 때문이었는데 다행히 추후에 양호한 냉각수 상태가 확인되어 우려할 상황을 피할 수 있었다. 본 논문에서는 후쿠시마 원전 사고를 계기로 국내 원자력 발전소내 핵연료 임시 저장시설의 안전성과 관련하여 중대사고 관점에서 검토해 보고자 한다.

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