• 해양환경안전학회지
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• 제27권2호
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• pp.355-362
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• 2021

L.P SCR의 촉매 반응을 위해 선박의 발전기용 4행정 디젤엔진의 배기가스 온도를 높게 설계 할 수밖에 없었다. 본 연구의 목적은 밸브개폐시기와 연료분사시기를 조정을 통한 배기가스의 온도 감소가 L.P SCR의 운전조건을 만족시키고 고온으로 인한 발전기 엔진의 사고를 예방하기 위함이었다. 배기가스 온도를 하강시키기 위해 캠샤프트의 각도를 조정하고 연료분사펌프의 Shim을 추가하였다. 그 결과 최대폭발압력은 12.8 bar 증가하였고 터보차저 출구온도 평균값은 13.3 ℃ 하강하였다. 터보차저 출구에서 SCR 입구까지의 열손실을 감안하더라도 L.P SCR 운전조건인 SCR 챔버 입구 온도인 290 ℃를 만족하였다. 배기가스 온도 하강을 통해 디젤발전기의 안전운전이 가능하게 한 연구였다.

• 전기화학회지
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• 제10권4호
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• pp.301-305
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• 2007

연료전지 운전 중에 스택(stack) 분리판 접착부위나 다른 경로로 부동액이 누설될 경우에는 화학적 반응에 영향을 주어 성능의 저하가 발생할 수 있다. 본 연구에서는 부동액이 누설되었을 경우의 성능 거동을 관찰하는 실험을 수행하였다. $400mA/cm^2$ 전류밀도 조건에서 마이크로 펌프를 이용하여 부동액을 주입하였으며 상대습도 100%/100%와 수소와 공기의 양론비는 1.5/2.0으로 고정하여 실험을 수행하였다. 3 cell stack을 이용하여 부동액을 주입한 후 정전류 회복 실험을 수행한 결과 cathode측에 부동액을 주입하였을 경우에는 성능이 회복되었고 anode측에 부동액을 주입하였을 경우에는 성능이 회복되기 어려운 것으로 나타났다. Anode측이 회복되지 못하는 이유로는 ethylene glycol의 산화반응에서 발생하는 불순물에 의한 피독 현상과 GDL과 3상 계면에 ethylene glycol이 물리적으로 흡착하였을 경우 반응에 필요한 연료 공급의 방해로 인한 성능 저하를 예상할 수 있다. 성능 저하에 영향을 주는 두 가지 변수를 확인하는 실험을 수행하였다. 회복 실험은 anode측에 water pump를 이용하여 질소 기체와 물을 동시에 공급하는 방법으로 실험을 수행하였고, 1시간 간격으로 성능 회복 유무를 확인하였다. 성능 평가는 polarization curve, cyclic voltammetry(CV), electrochemical impedance spectroscopy(EIS)를 사용하였으며, 정량분석은 gas chromatography를 이용하여 분석하였다. 부동액 주입 후 성능은 크게 저하되었고 정전류 회복 실험에서도 성능 회복은 미미하게 나타났다. 이 후 물 주입회복 실험을 수행하였고 회복 실험을 수행한 2시간 이후에는 93% 이상의 회복을 관찰할 수 있었다.

• 대한기계학회논문집A
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• 제36권11호
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• pp.1371-1379
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• 2012

본 논문에서는 직접분사 가솔린엔진 부품에 의해서 발생하는 진동에 대한 기여도를 분석하는 방법을 제시한다. 본 연구에서는 부분기여도함수를 적용하여 부품 상호간의 관련성에 대한 진동원을 규명 하는데 사용하였다. 직접분사 가솔린 엔진 부품의 진동원을 규명하는데 부분기여도함수 방법을 사용하기 위해서는 시스템의 모델링이 필요하며 본 연구에서는 진동 발생 경로를 2 입력과 단일 출력계로 시스템을 모델링하였다. 이 모델링을 증명 하기 위해서, 직접분사가솔린 엔진의 진동원인 고압펌프, 연료레일, 인젝터, 고압센서에 3 축 가속도계 센서로 각 부품의 진동을 측정했다. 이 모델링을 바탕으로 각각의 진동원에 대한 부분기여도 함수를 구했으며, 직접분사 부품들의 각각의 진동 기여도를 계산하였다. 부분기여도 함수를 바탕으로 한 모델링을 통해 각 부품들에서 발생되는 진동 출력 기여 값을 정량적으로 도출하였다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Journal of Radiation Protection and Research
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• 제28권2호
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• pp.127-135
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• 2003

The primary contributors to the total occupational radiation exposure in operating nuclear power plants are operation and maintenance activities doting refueling outages. The Advanced Power Reactor 1400 (APR1400) includes a number of design improvements and plans to utilize advanced maintenance methods and robotics to minimize the annual collective dose. The major radiation exposure reduction features implemented in APR1400 are a permanent refueling pool seal, quick opening transfer tube blind flange, improved hydrogen peroxide injection at shutdown, improved permanent steam generator work platforms, and more effective temporary shielding. The estimated average annual occupational radiation exposure for APR1400 based on the reference plant experience and an engineering judgment is determined to be in the order of 0.4 man-Sv, which is well within the design goal of 1 man-Sv. The basis of this average annual occupational radiation exposure estimation is an eighteen (18) month fuel cycle with maintenance performed to steam generators and reactor coolant pumps during refueling outage. The outage duration is assumed to be 28 days. The outage work is to be performed on a 24 hour per day basis, seven (7) days a week with overlapping twelve (12) hour work shifts. The occupational radiation exposure for APR1400 is also determined by an alternate method which consists of estimating radiation exposures expected for the major activities during the refueling outage. The major outage activities that cause the majority of the total radiation exposure during refueling outage such as fuel handling, reactor coolant pump maintenance, steam generator inspection and maintenance, reactor vessel head area maintenance, decontamination, and ICI & instrumentation maintenance activities are evaluated at a task level. The calculated value using this method is in close agreement with the value of 0.4 man-Sv, that has been determined based on the experience aid engineering judgement. Therefore, with the As Low As Reasonably Achievable (ALARA) advanced design features incorporated in the design, APR1400 design is to meet its design goal with sufficient margin, that is, more than a factor of two (2), if operated on art eighteen (18) month fuel cycle.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.