• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.357-362
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• 1996

LOFT LBLOCA test, LP-02-6 was analyzed using RELAP5/MOD3.2. It has a distinguished thermal-hydraulic phenomenon of a positive bottom-up core flow in tile blowdown phase. A modified nodalization which is based on that used in LP-LB-1 calculation by Lubbesmeyer was used in the calculation. RELAP5/MOD3.2 predicted overall system hydraulic behavior relatively well. However, the bottom-up quenching in the middle part of the core was not predicted sufficiently. It was demonstrated also that the peak cladding temperature can be predicted well by adjusting a discharge coefficient. But more improvements are needed in order to apply this code to actual plants with less user dependency.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.751-756
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• 1998

고리 1호기 원전 수명 연장을 위한 가압열충격(Pressurized Thermal Shock : PTS) 해석은 확률론적 안전성 평가 방법에 따라 수행된다. 본 연구는 가압열충격 상세 해석 연구의 일환으로 가압열충격 해석을 위한 계통해석시 사용되는 최적 평가(Best Estimate) 방법과 기존의 PCT(Peak Cladding Temperature) 관점의 해석에 사용되는 결정론적 안전성 평가 방법간의 해석 방법론 차이에 의한 열수력 거동의 상이점을 평가하기 위함이다. 이를 위해 1998년 설치 예정인 고리 1호기 교체 증기발생기(Replacement Steam Generator ; RSG) 안전성 분석 보고서$^{[1]}$ 의 주증기관 파단사고 해석 결과와 동일한 파단 크기 및 운전 출력에 대해 최적 평가 방법론에 따라 해석된 본 연구의 해석 결과를 비교, 평가하였다. 해석 결과 전출력 소형 주증기관 파단 사고에서는 터빈 유량 모델링 및 반응도 계수, 고온 영출력 대형 파단 사고에서는 가압기 모델, 반응도 계수 및 정지여유도가 해석 방법론에 따른 열수력 거동의 차이에 영향이 큰 것으로 평가되었다

• International Journal of Aeronautical and Space Sciences
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• 제17권1호
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• pp.37-44
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• 2016

Optical fiber temperature sensing systems have incomparable advantages over traditional electrical-cable-based monitoring systems. However, the fiber optic interrogators and sensors have often been rejected as a temperature monitoring technology in real-world industrial applications because of high cost and over-specification. This study proposes a multiplexed fiber optic temperature monitoring sensor system using an economical Optical Time-Domain Reflectometer (OTDR) and Hard-Polymer-Clad Fiber (HPCF). HPCF is a special optical fiber in which a hard polymer cladding made of fluoroacrylate acts as a protective coating for an inner silica core. An OTDR is an optical loss measurement system that provides optical loss and event distance measurement in real time. A temperature sensor array with the five sensor nodes at 10-m interval was economically and quickly made by locally stripping HPCF clad through photo-thermal and photo-chemical processes using a continuous/pulse hybrid-mode laser. The exposed cores created backscattering signals in the OTDR attenuation trace. It was demonstrated that the backscattering peaks were independently sensitive to temperature variation. Since the 1.5-mm-long exposed core showed a 5-m-wide backscattering peak, the OTDR with a spatial resolution of 40 mm allows for making a sensor node at every 5 m for independent multiplexing. The performance of the sensor node included an operating range of up to $120^{\circ}C$, a resolution of $0.59^{\circ}C$, and a temperature sensitivity of $-0.00967dB/^{\circ}C$. Temperature monitoring errors in the environment tests stood at $0.76^{\circ}C$ and $0.36^{\circ}C$ under the temperature variation of the unstrapped fiber region and the vibration of the sensor node. The small sensitivities to the environment and the economic feasibility of the highly multiplexed HPCF temperature monitoring sensor system will be important advantages for use as system-integrated temperature sensors.

• Nuclear Engineering and Technology
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• 제20권1호
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• pp.54-64
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• 1988

FLHT-2 실험 결과를 사용하여 원자력발전소의 중대사고발생시 노심의 거동을 해석하기 위한 전산코드인 SCDAP코드를 평가하였다. 계산결과에 의하면 코드는 실험시 측정된 노심의 온도경향, 총수소발생량 및 순간최대수소발생율, 그리고 연료봉내압과 피복재파열시간을 잘 예측하는 것으로 평가되었다. 그러나 이상유체높이와 복사열전달 및 zircaloy의 급격한 산화 시작 온도에 대한 모델은 수정되어야 할 것으로 평가되었다. 또한 핵 연료봉에서의 gap을 고려하여주는 것은 노심손상현상의 정확한 예측에 커다란 도움을 줄 수 있다는 것이 밝혀졌다.

• Nuclear Engineering and Technology
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• 제21권2호
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• pp.89-98
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• 1989

지르칼로이 -4 피복관에 대해서 3가지 변형 속도로(1.2$\times$10E-7/s, 2.0$\times$10E-6/s, 3.2$\times$10E -5/s), 553-873K의 온도 구간에서 구리 맨드렐 팽창 시험법을 공기와 진공(5$\times$10E-5 torr) 분위기에서 수행했고, 변형 속도의 변화는 시편의 가열 속도를 조절함으로써 얻을 수 있었다. 각각의 변형 속도에서 항복 응력 피크와 변형 속도 감도 최저값 그리고 활성화 부피 극대값이 나타나는 이유는 동적변형시효 현상 때문이라고 설명된다. 항복 응력 피크가 나타나는 온도와 변형속도로부터 얻어진 동적변형시효의 활성화 에너지는 196(KJ/mol)이었고 이 값은 $\alpha$-지르코니움과 지르칼로이-2에서 활성화 에너지(207-220 KJ/mol)값과 잘 일치한다. 그러므로 573-673K의 온도 구간에서 나타나는 동적변형시효 현상은 산소원자 때문이라고 생각된다. 산화에 의한 항복 응력의 증가는 공기중 실험과 진공 실험으로 얻어진 항복 응력값을 비교함으로서 얻었고, 그것은 항복 응력의 증가 분율로 표시했다. 결과는 변형속도가 느릴 수록 증가 분율은 더욱 더 커짐을 알 수 있었다. 그리고 산소 침투량과 항복 응력 증가 사이의 관계가 직선적이라는 가정하에 공기와 수중에서의 산화 속도를 비교하여 수중에서의 항복 응력 값을 계산해 보았다.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.541-546
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• 2014

Probabilistic Safety Assessment (PSA) has been widely used to estimate the overall safety of nuclear power plants (NPP) and it provides base information for risk informed application (RIA) and risk informed regulation (RIR). For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs) for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT) were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2847-2858
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• 2021

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.429-434
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• 1996

K-REM［1］, which is under development as a realistic evaluation model of large break LOCA, is applied to the analysis of cold leg guillotine break of Ulchin 3&4. Fuel parameters on which statistical analysis of their effects on the peak cladding temperature (PCT) are made and system parameters on which the concept of limiting value approach (LVA) are applied, are determined from the single parameter sensitivity study. 3 parameters of fuel gap conductance, fuel thermal conductivity and power peaking factor are selected as fuel related ones and 4 parameters of axial power shape, reactor power, decay heat and the gas pressure of safety injection tank (SIT) are selected as plant system related ones. Response surface of PCT is generated from the plant calculation results and on which Monte Carlo sampling is made to get plant application uncertainty which is statistically combined with code uncertainty to produce the 95th percentile PCT. From the break spectrum analysis, blowdown PCT of 1350.23 K and reflood PCT of 1195.56 K are obtained for break discharge coefficients of 0.8 and 0.5, respectively.

• Nuclear Engineering and Technology
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• 제37권4호
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.