• Nuclear Engineering and Technology
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• 제32권1호
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Nuclear Engineering and Technology
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• 제42권6호
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• pp.690-703
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• 2010

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between $5^{th}$ and $95^{th}$ percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure.

• 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.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.54-70
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• 2017

The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility-the advanced core-cooling mechanism experiment (ACME)-was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups-a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break-were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative effects on the passive core cooling performance caused by nitrogen injection during the SBLOCA transient.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• 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.

• 한국음향학회지
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• 제28권2호
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• pp.155-163
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• 2009

Text-to-speech 시스템에서 입력 텍스트로부터 운율 정보를 생성하기 위해서는 운율구 경계, 음소 지속시간, 기본주파수 포락선 설정의 3가지 기본적인 모듈이 필요하다. Break 인덱스 (BI; Break Index)는 합성기에서 운율구의 경계를 나타내고, 자연스러운 합성음을 생성하기 위해서는 BI를 정확히 예측하여야 한다. 그러나 BI는 문장의 의미나 화자의 읽기 습관(reading style)에 따라 임의적으로 결정되는 경우가 많아 정확한 예측이 매우 어렵다. 특히 일본어 합성기에서는 악센트 구 경계 (APB; Accentual Phrase Boundary)와 major phrase 경계 (MPB; Major Phrase Boundary)의 정확한 예측이 어렵다. 따라서 본 논문에서는 APB와 MPB 예측 오류를 보완할 수 있는 방법을 제안한다. BI를 고정 break (FB; Fixed Break)와 가변 break (VB; Variable Break)로 분류하여 합성단위 선택을 수행한다. 일반적으로 BI는 한번 생성되면 변하지 않는다. 따라서 BI가 잘못 생성된 경우 최적의 합성음을 생성할 수 없게 되는데, VB는 생성된 BI와 그것과 유사한 BI를 함께 이용하여 합성단위 선택을 수행함으로써 합성음의 BI가 생성된 BI와 다를 수 있는 것을 의미한다. APB와 MPB에 해당하는 BI에 대하여 VB인지 FB인지 CART(Classification and Regression Tree)를 이용하여 예측하고, VB인 경우 기본 주파수와 음소 지속시간에 대해 다중 운율 모델을 생성하여 합성단위 선택을 수행하였다. MOS 테스트 결과 원음이 4.99, 제안한 방법을 4.25, 기존의 방법은 4.01로 합성음의 자연성을 향상시킬 수 있었다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2002년도 추계학술대회 논문집 전기물성,응용부문
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• pp.179-182
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• 2002

XLPE compound have used for insulation of 22.9(kV) power cable. But tree retardant power cable has developed and is going to br used commonly. TR XLPE compound retard production and growth of water tree. In this paper, tensile strength, elongation at break, degree of crosslinking, lightning impulse test, AC breakdown test, cyclic aging for 14days and accelerated water treeing test of TR XLPE insulated power cable were examined according to the KEPCO buying spec. & AEIC CS 5-94 standards. before and after As the result, tensile strength, elongation at break and degree of crosslinking test results of TR XLPE insulation were higher than requirement values. After accelerated water treeing test for 120 days, 240 days and 360 days, AC breakdown voltages were not decreased for accelerated water treeing aging duration

• Tribology and Lubricants
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• 제25권2호
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• pp.102-107
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• 2009

The thermal deformation volume of motorcycle break disk was studied using a disk-on-pad type friction tester. Thermal deformation volume of motorcycle break disk have an effect on the frictional factor such as applied load, sliding speed, sliding distance and number of ventilated disk hole. However, it is difficult to know the mutual relation of these factors on thermal deformation volume. In this study, the thermal deformation volume with ANSYS workbench are obtained by application of temperature from mechanical test. From this study, the result was shown that the motorcycle break disk with ventilated hole 3 have the most excellent thermal deformation characteristics. The regression equation with frictional factors which have a trust rate of 95% for prediction of thermal deformation volume of motorcycle break disk was composed.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.108-127
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• 2000

Hot leg break LBLOCA(Large Break LOCA) had a potential to be a containment maximum pressure accident in YGN3&4, which was induced from excessive conservatism in the CE analysis methodology of mass and energy release. This study conducted mass and energy release experiment for the hot leg break LBLOCA during post blowdown with an integral test facility, SNUF(Seoul National University Facility). This facility simulated YGN 3&4 with volume ratio of 1/1140 based on Ishii's three level scaling. Experiment showed that SI(Safety Injection) water refilled cold leg first and core later. SI water was vaporized in the core, which resulted in the repressurization of reactor. This increase of pressure drove the water in cold leg to flow up half height of U tubes. However, since the water was drained back soon, the release through the SG side broken section by evaporation was negligibly small. This study also provided experimental assessment of RELAP5 results by KAERI for the release through the SG side broken section.