• Nuclear Engineering and Technology
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• 제51권2호
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• 한국안전학회지
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• 제36권2호
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Elastomers and Composites
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• 제56권3호
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• pp.124-135
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• 2021

Evaluating service lives of rubber materials at certain temperatures requires a destructive method (typically using elongation at break). In this study, a non-destructive method based on hardness change rate was proposed for evaluating the service life of chloroprene rubber (CR). Compared to the destructive method, this non-destructive method ensures homogeneity of CR specimens and requires a small number of samples. Thermal accelerated degradation test was conducted on the CR specimens at 55, 70, 85, 100, and 125℃, and the tensile strength, elongation at break, and hardness were measured. The results of the experiment were compared to those of the accelerated life evaluation method proposed in this study. Comparing the analyzed lives in the high temperature region (70, 85, 100, and 125℃), the difference between the service lives for the destructive method (using the elongation at break) and non-destructive method (using the hardness) was approximately 0.1 year. Therefore, it was confirmed that the proposed non-destructive evaluation method based on hardness changes can evaluate the actual life of CR under thermally accelerated degradation conditions.

• 동력기계공학회지
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• 제3권3호
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• pp.22-28
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• 1999

The injected liquid does not break-up instantly after injection for diesel engine. There is some unbroken portion, which is the liquid core(The length of liquid core is called the break-up length) in the spray. If the liquid core is longer than the depth of the bowl in the small DI diesel engine, the liquid core impinges on the surface of the piston. Once the liquid core impinges on the surface, it cannot ignite or burn rapidly and thus prolongs burning time with a degradation in thermal efficiency. The break-up length of a diesel spray in a compressure vessel was measured by an electric resistance method, A voltage was applied between the nozzle and screen, bar, needle electrode inserted at various axial and radial positions into atomizing sprays. As a result, a current flows not only in the region of liquid core but also through the droplets of the spray. It is found that the break-up length measured with screen electrode is overestimated. The break-up length of the spray is found to be proportional to the square root of the density ratio of fuel and surrounding gas. The break-up length of the spray decreases as the injection pressure and the back pressure increase.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.719-726
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• 2012

The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.

• 대한기계학회논문집B
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• 제40권3호
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• pp.165-172
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• 2016

노심보충탱크(Core Makeup Tank, CMT), 안전주입탱크(SafetyInjection Tank, SIT)와 자동감압계통(Auto Depressurization System, ADS)로 구성된 1 계열의 SMART 피동안전주입계통의 주입특성을 파악하기 위한 소형냉각재상실사고(SBLOCA) 모의에 대한 실험적 연구가 수행되었다. SBLOCA의시험은 0.4 인치 안전주입수 배관파단에 대해 수행되었으며, 정상상태 조건은 실험요건서에 제시된 시험 초기 조건을 만족시키도록 746초 동안 운전되었다. 노심 출력 및 안전주입 유량 등의 경계 조건도 적절히 모의되었으며, 안전주입계통 배관에서의 파단, 히터 트립 및 잔열곡선 인가, 원자로냉각재펌프 관성서행(Coastdown), 급수 중단, CMT 및 SIT의 주입, ADS #1 개방이 SBLOCA 시나리오에 따라 적절히 모의되었다. 노심지지원통 내부의 액체환산수위는 파단 초반에 감소하다가 CMT와 SIT가 주입되면서 서서히 회복되었으며, 피동안전주입계통의 주입유량이 노심 수위를 회복하기에 충분한 것으로 판단할 수 있다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.641-642
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• 2005

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.228-233
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• 2000

The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

• 한국안전학회지
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• 제16권4호
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• pp.29-38
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• 2001

The purpose of this paper is to evaluate the structural integrity of a reactor pressure vessel subjected to the pressurized thermal shock(PTS) during the transient events, such as main steam line break(MSLB) and small break loss of coolant accident(SBLOCA). For postulated surface or subsurface cracks, variation curves of stress intensity factor are obtained by using the three different methods, including ASME section XI code anlysis, the finite element alternating method and the finite element method. From the stress intensity factor curves, the maximum allowable nil-ductility transition temperatures(RT/NDT/) are determined by the tangent criterion and the maximum criterion for various crack configurations and two initial transient events. As a result of the analysis, it is noted that axial cracks have smaller maximum allowable RT$_{NDT}$ values than same-sized circumferential cracks for both the transient events in the case of the tangent criterion. Axial cracks have smaller RT$_{NDT}$ values than same-sized circumferential cracks for MSLB and circumferential cracks have smaller values than axial cracks for SBLOCA in the case of the maximum criterion.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.702-708
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• 2019

This paper deals with break size estimation of loss of coolant accidents (LOCA) using a nonlinear autoregressive with exogenous inputs (NARX) neural network. Previous studies used static approaches, requiring time-integrated parameters and independent firing algorithms. NARX neural network is able to directly deal with time-dependent signals for dynamic estimation of break sizes in real-time. The case studied is a LOCA in the primary system of Bushehr nuclear power plant (NPP). In this study, number of hidden layers, neurons, feedbacks, inputs, and training duration of transients are selected by performing parametric studies to determine the network architecture with minimum error. The developed NARX neural network is trained by error back propagation algorithm with different break sizes, covering 5% -100% of main coolant pipeline area. This database of LOCA scenarios is developed using RELAP5 thermal-hydraulic code. The results are satisfactory and indicate feasibility of implementing NARX neural network for break size estimation in NPPs. It is able to find a general solution for break size estimation problem in real-time, using a limited number of training data sets. This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr NPP.