• Nuclear Engineering and Technology
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• 제31권4호
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• pp.375-384
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• 1999

The credibility of CHECWORKS FAC model analysis was evaluated for plant application in a model plant chosen for demonstration. The operation condition at each pipe component was defined before the wear rate analysis by plant data base, water chemistry analysis, and network flow analysis. The predicted wear was compared with the measured wear for 57 sample components selected from 43 susceptible line groups analysed. The inspected 57 locations represent components of highest predicted wear in each line group. Both absolute value and relative ranking comparisons indicated reasonable correlations between the predicted and the measured values. Four components showed much higher measured wear rates than the predicted ones in the feed water train from main feed water pump discharge to steam generator, probably due to high hydrazine concentration operation the effect of which had not been incorporated into the CHECWORKS model. The measured wear was higher than the predicted one consistently for components with least susceptibility to FAC. It is believed that the conservatism maintained during UT data analysis dominated the measurement accuracy. A great deal of enhancement is anticipated over the current plant pipe management program when a comprehensive plant pipe management program is implemented based on the model analysis.

• 태양에너지
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• 제1권1호
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• pp.1-11
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• 1981

The electric generation system by solar energy was built which is composed of $10m^2$ reflector, parabolic mirror and the absorbers. The absorber(I) is a single iron pipe and the absorber (II) contains seven small iron pipes. The ratio of the area of the reflectors to that of the absorber is around 99.4-440. The absorber(II) is more efficient in power than (II) by 5.6 percent. The steam power efficiency of the absorber (II) is 25 percent in this experiments and 20 percent efficiency would be expected for 80.000 Kilowatts.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2013년도 추계학술대회 논문집
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• pp.868-869
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• 2013

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1997년도 추계학술대회논문집; 한국과학기술회관; 6 Nov. 1997
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• pp.271-278
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• 1997

The flow of steam through a safety valve vent pipe system in the boiler has been analyzed to provide a design basis of diffuser silencer for attenuating shock-shell and jet noise. Numerical analysis to estimate inner fluid of silencer and noise propagation outside silencer are performed. The distribution curve of fluid information to provide average values about inner fluid of silencer is presented by theoretical analysis.

• 대한기계학회논문집B
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• 제41권5호
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• pp.353-363
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• 2017

부하변동과 외란에 따른 관류보일러 물-증기 계통의 동적 거동을 물리적 모델링 접근방법으로 모사하였다. 본 논문에서는 수관의 질량, 에너지와 운동량 방정식을 고려한 아임계 영역의 동적 모사를 보고한다. 500MWe 급의 절탄기, 증발기와 과열기로 이루어진 단순한 보일러 시스템을 가정하였고, 증발기 모델링은 참고문헌 데이터와 검증을 진행 하였다. 이 시스템에 대하여 외란에 따른 정량적 응답특성을 살펴보았다. 또한, 수연비(증기량과 연료의 유량비)가 설계조건과 크게 다른 탈 설계 운전 사례에 대한 보일러 시스템의 동적 응답평가를 진행 하였다. 그 결과를 통해 적절한 수연비의 제어와 재순환 시스템과 분무 감온기 설계의 중요성이 재확인 되었다.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• 한국전산유체공학회지
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• 제20권1호
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• pp.39-46
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• 2015

In this research, we studied the pressure drop affecting on the internal surface roughness and the curvature radius of a U-tube, which is used for the cooling system in PWR(Pressurized Water Reactor). Using ANSYS-FLUENT, a commercial code based on CFD(Computational Fluid Dynamics) technique, we compared a Moody chart with the Darcy friction factor changed by a range of various surface roughness and Reynolds numbers of a straight pipe model. We studied the effect giving variation about a range of various surface roughness and the curvature radius of the full scale U-tube model. The material of the heat transfer tube is Inconel 690 used in the steam generator. We compared the velocity distribution of selected 4 locations, and derived the correlation between the surface roughness and the pressure drop for the U-tube of each representative curvature radius using the linear regression method.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 춘계학술대회논문집A
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• pp.37-42
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• 2000

The boiler tubes and steam Pipes operating both at high temperature and pressure for a long period of time in a power plant are degraded by creep because of internal pressure. So, the remaining life of a component is evaluated by the creep rupture strength. Although the conventional method to evaluate the creep damage is widely used, it has some disadvantages such as requires large size specimen and long employed to evaluate the correlation between fracture toughness and evaluation time. Recently, new method so called "small lunch test' is used to evaluate degradation of creep. In this study, a conventional creep test and a small punch test are conducted using 2.25Cr-1Mo steel which is mainly used for the boiler tubes and steam pipes in power plant. The creep life, approximately 1,500 hrs, is determined by conventional method under a severe condition then specimens for a small Punch test are obtained after certain time intervals such as 1/4, 1/2 and 3/4 of final rupture time, respectively.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.51-59
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• 2017

Steam jets ejected from a rupture zone of high energy pipes may cause damage to adjacent structures. This event could lead to more serious accidents in nuclear power plants. Therefore, to prevent serious accidents, high energy pipes of nuclear power plants are designed according to the ANSI / ANS 58.2 technical standard. However, the US Nuclear Regulatory Commission (USNRC) has recently pointed out non-conservatism in existing high energy pipe fracture evaluation methods, and required the assessment of the unsteady load of the jet caused by a potential feedback mechanism as well as the impact range of steam jet, the jet impact loads and the blast wave effects at the initial breakage stage. The potential feedback mechanism refers to a phenomenon in which a vortex formed by impingement jets amplifies vortex itself and induces jet vibration in a shear layer. In this study, CFD methodology using the LES turbulence model is established and numerical analysis is carried out to evaluate the dynamic behavior of impingement jets and the potential feedback mechanism during jet impingement. Obtained results have been compared with an empirical correlation and experiment.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.797-802
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• 2014

This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO) was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.