• 방사성폐기물학회지
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• 제18권4호
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• pp.481-496
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• 2020

After nuclear power plants are permanently shut down and decommissioned, the remaining irradiated metal components such as stainless steel, carbon steel, and Inconel can be used as neutron absorber. This study investigates the possibility of reusing these metal components as neutron absorber materials, that is burnable poison. The absorption cross section of the irradiated metals did not lose their chemical properties and performance even if they were irradiated over 40-50 years in the NPPs. To examine the absorption capability of the waste metals, the lattice calculations of WH 17×17 fuel assembly were analyzed. From the results, Inconel-718 significantly hold-down fuel assembly excess reactivity compared to stainless steel 304 and carbon steel because Inconel-718 contains a small amount of boron nuclide. From the results, a 20wt% impurity of boron in irradiated Inconel-718 enhances the excess reactivity suppression. The application of irradiated Inconel-718 as a burnable absorber for SMR core was investigated. The irradiated Inconel-718 impurity with 20wt% of boron content can maintain and suppress the whole core reactivity. We emphasize that the irradiated metal components can be used as burnable absorber materials to control the reactivity of commercial reactor power and small modular reactors.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1135-1144
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• 2024

With the increasing ratio of renewables in the grid, a low-carbon and stable base load source that also is capable of load tracking is in demand. Sodium cooled fast reactors (SFRs) coupled to thermal heat storage system (THS) is a strong candidate for the need. This research focuses on the designing and performance validation of a two-tank THS based on molten salt to integrate with a 280 MW_th sodium cooled fast reactor. Designing of the THS includes the vital component, sodium-to-salt heat exchanger which is a technology gap that needs to be filled, and designing and parameter selection of the tanks and related pumps. Modeling of the designed THS is conducted followed by the description of operation strategies and control logics of the THS. Finally, the dynamic simulation of the designed THS is conducted based on Fortran. Results show, the proposed power system meets the need of the design requirements to store heat for 18 h during a day and provide 500 MW_th for peak demand for the rest of the day.

• 응용통계연구
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• 제29권7호
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• pp.1385-1398
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• 2016

원자력 발전소의 가압기는 1차 계통의 냉각재가 고온에서도 기화되지 않도록 압력을 가해주는 장치이다. 즉, 가압기의 고장은 원자력 발전소에 큰 영향을 미칠 수 있으며, 따라서, 가압기의 조기 고장 감지는 원자력 발전소의 안전에 매우 중요하다. 이를 위해, 본 연구에서는 마할라노비스 거리 개념과 다구찌 품질 공학 이론에 기반한 패턴 분류 인식 알고리즘 중 하나인 마할라노비스 다구찌 시스템(MTS)을 가압기 압력 제어 계통의 조기 고장 감지에 적용하였다. MTS의 고장 감지 성능을 검증하기 위해, 실제 원자력 발전소에서 발생하고 있는 가압기 압력전송기 고장 시나리오를 대상으로 하여, Full Scope 시뮬레이터를 통해 모사된 데이터에 적용하였다. 실험 결과, MTS는 단일 센서모니터링 기반의 전통적인 고장 감지에 비하여 매우 빠르게 고장을 감지할 수 있었다.

• Nuclear Engineering and Technology
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• 제14권4호
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• pp.196-203
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• 1982

본 논문에서는 고리 1호기의 기사용 핵연료 집합체를 수송하기 위한 Cask를 설계하였다. 이를 위하여 고리 1호기의 기사용 핵연료 집합체로부터 방출되는 감마선과 중성자를 계산하여 MORSE 및 ANISN전산 코드로써 차폐 계산을 수행하였다. 그 결과, 9개의 집합체를 동시에 수송할 수 있는 Steel Cask가 가장 적합하다는 것을 밝혔다. 이 Steel Cask에 대한 안전성을 평가하기 위하여 연료봉의 중심 온도와 복재온도를 계산하여 핵연료의 용융점보다 훨씬 낮음을 증명하였다. 또한 KENO와 MORSE전산 코드를 사용하여 임계도 계산을 수행하여 미임계 상태임을 증명하였다. 이로써 9개의 기사용 핵연료 집합체를 동시에 수송할 수 있는 Steel Cask를 간단히 설계하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.178-183
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• 2004

For developing fatigue design curve of cast stainless steels that would be used in piping material of domestic nuclear power plants, a low-cycle fatigue test rig was built. It is capable of performing tests in pressurized high temperature water environment of PWR. Cylindrical specimens of CF8M were used for the strain-controlled environmental fatigue tests. Fatigue life was measured in terms of the number of cycles with the variation of strain amplitude at 0.04%/s strain rates. The disparity between target length and measured length of specimens was corrected by using finite element method. The corrected test results showed similar fatigue life trend with another previous results.

• 에너지공학
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• 제8권1호
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• pp.110-118
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• 1999

In order to protect the nuclear reactor coolant system from corrosion, lithium is injected into the coolant from the chemical injection tank. The present study investigates the chemical injection characteristics of the injection tank using a low Reynolds number turbulence model. Laminar flow analysis showed very little diffusion of the jet and gave incorrect flow and concentration fields. A disk located near the inlet of the injection tank was effective in mixing the chemical additives in the top portion of the tank, and significant reduction in injection time was obtained.

• Journal of Welding and Joining
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• 제10권4호
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• pp.240-249
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• 1992

The objective of this study is to investigate the characteristics of welded joints on the Zircaloy-4 resistance upset welding for HWR(Heavy Water reactor)fuel rods. To estimate the characteristics of welded joints, the various tests were performed on the test coupons systematically with a wide range of each welding parameters in terms of a tensile test, burst test, knoop hardness test and metallography. Major results obtained in this study are as follows: 1. The tube and machined with 120.deg. projection was the reliable weld joint design for the nuclear fuel rod end cap welding. 2. As the weld current and the amount of upset increased linearly with increasing welding main heat input, it could make an estimate of their variation in accordance with the phase shift control. 3. It was found that an increase in squeeze force has an effect on the upset contour of welded joint because the amount of upset were increased by the change of squeeze force.

• 대한기계학회논문집A
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• 제26권9호
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• pp.1897-1904
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• 2002

Monotonic torsional and pure torsional low cycle fatigue(LCF) test with artificial degradation were performed on duplex stainless steel(CF8M). CF8M is used in pipes and valves in nuclear reactor coolant system. It was aged at 430$^{\circ}C$ for 3600hrs. Through the monotonic and LCF test, it is found that mechanical properties(i.e., yield strength, strain hardening exponent, strength coefficient etc.) increase and fatigue life(N$\sub$f/) decreases with degradation of material. The relationship between shear strain amplitude(${\gamma}$$\sub$a/)and N$\sub$f/ was proposed.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1993년도 추계학술발표회 초록집
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• pp.103-108
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• 1993

In nuclear power plants using pressurized water as the main coolant, it is necessary to maintain system pressure within operational range. During transients, the coolant shrinks and expands causing insurge and outsurge of coolant in the pressurizer. In CANDU system, the pressure is controlled mainly by the pressurizer/degasser-condenser system. In CANDU system, the control of heat transport system pressure is achieved by giving heat to the pressurizer by activating the heaters to compensate a diminution in pressure or by removing heat from the pressurizer by bleeding steam to the degasser-condenser to compensate an increase in pressure. This study aims at developing a theoretical model capable to simulate various operational transients in the CANDU primary heat transport system (PHTS), applicable to CANDU engineering simulator on real time basis.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1992년도 봄 학술발표회 논문집
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• pp.107-113
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• 1992

The heat of hydration of cement the causes the internal temperature rise at early age, particulary in massive concrete structures such as a mat-slab of nuclear reactor building or a dam or a large footing. As the result of the temperature rise and restraint of foundation, the thermal stress enough to induce concrete cracks can occur. Therefore, the prediction of the thermal stress is very important in the design and construction stages in order to control the cracks developed in massive concrete structures. And, more creep and shrinkage take place at elevated temperatures in young concrete, Thus the effect of creep and shrinkage must be considered for checking the safety and servicebility(crack, durability and leakage).