• Nuclear Engineering and Technology
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• 제53권3호
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• pp.1029-1040
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• 2021

For the structural integrity evaluation of pressurized water reactor (PWR) steam generator (SG) tubes subjected to transient hydraulic loading, determination of the tube-to-tube gap velocity and static pressure distributions along the tubes is prerequisite. This paper addresses both computational fluid dynamics (CFD) and analytical approaches for predicting the tube-to-tube gap velocity and static pressure distributions during blowdown following a feedwater line break (FWLB) accident at a PWR SG. First of all, a comparative study on CFD calculations of the transient velocity and pressure distributions in the SG secondary sides for two different models having 30 or no tubes is performed. The result shows that the velocities of sub-cooled water flowing between any adjacent two tubes of a tubed SG model during blowdown can be roughly estimated by applying the specified SG secondary side porosity to those of the no-tubed SG model. Secondly, simplified analytical approximate solutions for the steady two-dimensional SG secondary flow velocity and pressure distributions under a given discharge flowrate are derived using a line sink model. The simplified analytical solutions are validated by comparing them to the CFD calculations.

• 한국초전도ㆍ저온공학회논문지
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• 제10권1호
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• pp.57-61
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• 2008

A 1MW class HTS(High-Temperature Superconducting) synchronous motor has been developed. This motor was aimed to be utilized for industrial application such as large motors operating in large plants. The HTS field windings of the developed motor is cooled by way of Neon thermosiphon mechanism and the stator coil is cooled by water through hollow copper conductor. This paper describes performance test results of our motor, which was conducted at steady state in generator mode and motor mode.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.93-100
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• 2016

Since the sodium-cooled fast reactor is operated in a hostile environment due to the use of liquid sodium as its coolant, advanced techniques for in-service inspection are required to periodically verify the integrity of the reactor. This paper presents the development of in-service inspection techniques for Proto-type Generation IV Sodium-cooled Fast Reactor. First, the 10 m long plate-type ultrasonic waveguide sensor has been developed for in-service inspection of reactor internals, and its feasibility was verified through several under-water and under-sodium experiments. Second, the combined inspection system for in-service inspection of ferromagnetic steam generator tubes has been developed. The remote field eddy current testing and magnetic flux leakage testing can be conducted simultaneously by using the developed inspection system, and the detectability was demonstrated through several damage detection experiments. Finally, the electro-magnetic acoustic transducer which can withstand high temperature and be installable in the remote operated vehicle has been developed for in-service inspection of the reactor vessel, and its detectability was investigated through damage detection experiments.

• 설비공학논문집
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• 제16권5호
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• pp.438-443
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• 2004

The objective of this paper is to study the effects of the input gas flow rate and the ambient temperature variation on the absorption cycle performance. An air-cooled NH$_3$-$H_2O$ absorption chiller is tested in the present study. The nominal cooling capacity of the single effect maching is 17.6 ㎾ (5.0 USRT). The cooling capacity, coefficient of performance, burner efficiency, and each state point are measured with the variations of the heat input and the ambient temperature. It is found that the COP and cooling capacity increase with increasing the generator exit temperature up to a certain temperature and then decrease. It is also found that the COP and the cooling capacity decrease with increasing the ambient temperature. The maximum COP of 0.51 is obtained from the present experiment.

• 한국항행학회논문지
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• 제20권6호
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• pp.556-561
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• 2016

논문은 ISM 밴드 응용분야에 사용되는 13.35 MHz, 5 kW RF 제너레이터의 구현에 대한 내용이다. 이 제너레이터의 구성은 드라이브증폭기와 LDMOS 1.25 kW 트랜지스터를 클래스 AB 푸쉬풀 증폭기로 구현한 모듈을 바탕으로 4개의 모듈을 윌킨슨 형태의 전송선로 변환기를 이용하여 결합하는 구조를 갖고 있다. 이 제너레이터는 선형성보다는 높은 출력과 효율을 얻는 것이 중요하며, 트랜지스터에서 발생하는 열은 워터 쿨링에 의한 수냉식 방법을 선택하여 외부로 배출하였다. 또한, 결합기와 저역통과 필터의 결합 그리고 출력 및 반사 신호를 검출하는 보호회로를 삽입하였다. 이렇게 구현된 RF 제너레이터는 포화 전력레벨 5.33 kW 출력에서 79 %의 효율을 얻었다.

• 한국추진공학회지
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• 제8권2호
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• pp.10-17
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• 2004

본 논문에서는 액체로켓엔진용 160 kW급 터보펌프의 터빈을 구동하고, 액체산소와 케로신을 추진제로 사용하는 연료 과잉 가스발생기의 설계점 연소성능시험 결과에 대해 논의하였다. 충돌형 F-O-F 분사기로 구성된 헤드부, 물냉각 채널 연소실, torch igniter, turbulence ring 그리고 측정 링을 갖는 가스발생기에 대해 기술하였고, 설계점에서의 연소시험 및 turbulence ring 장착여부. 연소실 길이 변화에 따른 연소시험의 결과들에 대해 기술하였다. 연소시험 결과 가스발생기는 설계점에서 안정된 작동성을 보여주었고. 연소압력 및 온도 등의 성능은 예측치에 근접하는 결과였다. Turbulence ring은 출구에서의 가스온도를 균일하게 분포시켜 효과적인 혼합 장치임을 보여 주었고, 4∼6msec 정도에서의 연소가스 잔류시간은 연소효율에 큰 영향을 주지 않았다. 가스발생기 출구에서의 온도는 공급되는 추진제의 O/F ratio에 따라 매우 민감하게 변화하였다.

• 대한기계학회논문집A
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• 제39권6호
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• pp.625-629
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• 2015

본 연구에서는 수냉식 발전기 고정자 권선의 건전성 예지 방법에 대해 연구하였다. 권선의 데이터를 흡습 데이터와 정상 데이터로 분류 하였으며 각각의 데이터 군을 다른 방법으로 예측 하였다. 흡습 데이터를 예측하기 위해 픽의 제 2 법칙(Fick's second law)를 이용하여 건전성 감소 모델링을 하였고 픽의 제 2법칙의 해를 이용하여 흡습 모델식을 만들었다. 정상 데이터는 데이터의 분포가 정규분포를 따른다는 가설을 세운 후 카이제곱 검정을 통해 이를 입증하였다. 예측된 흡습 데이터와 정상 데이터를 이용하여 건전성 인자인 방향성 마할라노비스 거리(Directional mahalanobis distance; DMD)의 예측값을 산출하였고 흡습 권선의 고장 예상시점을 계산했다.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.700-711
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• 2015

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called "self-wastage phenomena." A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.3122-3125
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• 2021

Electromagnetic vortex flowmeter is a new type of instrument for detecting leakage of steam generator, and the signal processing method based on the envelope to calculate energy ratio can effectively detect bubbles in sodium flow. The signal processing method is not affected by changes in the amplitude of the sensor output signal, which is caused by changes in magnetic field strength and other factors. However, the detection sensitivity of the electromagnetic vortex flowmeter is reduced. To this end, a signal processing method based on inverse Fourier transform to calculate energy ratio is proposed. According to the difference between the frequency band of the bubble noise signal and the flow signal, only the amplitude in the frequency band of the flow signal is retained in the frequency domain, and then the flow signal is obtained by the inverse Fourier transform method, thereby calculating the energy ratio. Using this method to process the experimental data, the results show that it can detect 0.1 g/s leak rate of water in the steam generator, and its performance is significantly better than that of the signal processing method based on the envelope to calculate energy ratio.