• Nuclear Engineering and Technology
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• 제38권6호
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• pp.585-590
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• 2006

This paper describes the results of an irradiation test and the specifications of the pneumatic transfer system (PTS) in the NAA #3 irradiation hole at the HANARO research reactor, which was reinstalled after some modifications of the operation mode at the end of 2004. The outer and inner diameters of the PE transfer tube are 34.1 and 27.5 mm, respectively. PE rabbit was used for sample irradiation. The $N_2$ gas pressure of the PTS lines was adjusted to 0.75 bar. The average sending time to the reactor was $8.5{\pm}0.3$ s and the average receiving time back to the receiver was $3.2{\pm}0.2$ s. The internal and external temperature of the irradiation tube was measured in a range of 50 to $80^{\circ}C$ for a 40 s to 80 s irradiation time, respectively. The optimum irradiation time was estimated to be less than 80 s. The thermal, epithermal and fast neutron flux at 30 MW thermal power were $1.42{\pm}0.01{\times}10^{14},\;1.51{\pm}0.04{\times}10^{13}$ and $9.48{\pm}0.69{\times}10^{11} n{\cdot}cm^{-2}{\codt}s^{1-}$, respectively. The cadmium ratio was approximately 9.40. The data obtained will be applied to supplement user information and for reactor management.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.323-331
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• 2015

Zirconium alloys are widely used in nuclear reactors as structural materials. During the operation, they are exposed to fast neutrons. Ion irradiation is used to simulate the damage introduced by neutron irradiation. In this article, we briefly review the neutron irradiation damage of zirconium alloys, then summarize the effect of ion irradiation on microstructural evolution, mechanical and corrosion properties, and their relationships. The microstructure components consist of dislocation loops, second phase precipitates, and gas bubbles. The microstructure parameters are also included such as domain size and microstrain determined by X-ray diffraction and the S-parameter determined by positron annihilation. Understanding the relationships of microstructure and properties is necessary for developing new advanced materials with higher irradiation tolerance.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.3996-4001
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• 2023

The High-flux Advanced Neutron Application Reactor (HANARO) produces radioisotopes (RIs) (¹³¹I, ¹⁹²Ir, etc.) through neutron irradiation on various RI production targets. Among them, ¹⁷⁷Lu and ¹⁶⁶Ho are particularly promising owing to their theranostic characteristics that facilitate simultaneous diagnosis and treatment. Prior to neutron irradiation, evaluating the nuclear heating of the RI production target is essential for ensuring the thermal-hydraulic safety of HANARO. In this study, the feasibility of producing ¹⁷⁷Lu and ¹⁶⁶Ho using irradiation holes of HANARO was investigated in terms of thermal-hydraulic safety. The nuclear heating rates of the RI production target by prompt and delayed radiation were calculated using MCNP6. The calculated nuclear heating rates were used as an input parameter in COMSOL Multiphysics to obtain the temperature distribution in an irradiation hole. The degree of temperature increase of the ¹⁷⁷Lu and ¹⁶⁶Ho production targets satisfied the safety criteria of HANARO. The nuclear heating rates and temperature distribution obtained through the in silico study are expected to provide valuable insight into the production of ¹⁷⁷Lu and ¹⁶⁶Ho using HANARO.

• 대한기계학회논문집A
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• 제26권1호
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• pp.188-195
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• 2002

To investigate the degradation of mechanical properties induced mainly by neutron irradiation, the tensile tests were conducted from room temperature to 300\\`c using the irradiated and the unirradiated Zr-2.5Nb pressure tube materials. The irradiated longitudinal and transverse specimens were collected from the coolant inlet, middle, and outlet parts of M-11 tube which had been operated in Wolsung CANDU Unit-1 and exposed to different operating temperatures and irradiation fluences. The different tensile behavior was characterized not by the fluences of irradiation but by the tensile loading direction. The transverse specimen showed the higher strength and lower elongation than those of the longitudinal one. It was believed that these phenomena resulted from the microstructure anisotropy caused by the extrusion process. The increased strength hardening and decreased elongation embrittlement of the irradiated material were compard to those of the unirradiated one. While the tensile strength of the inlet was higher than that of the outlet, the elongation of the inlet was lower than that of outlet. Considering the operation condition, it was proposed that the operating temperature could be a more effective parameter than the irradiation fluence for long-time life. Through the TEM observation, it was found that while the a-type dislocation density was increased, the c-type dislocation was not changed in the irradiated. The fact that the higher dislocation density was sequentially distributed over the inlet, the middle, and the outlet parts was consistent with the distribution of the tensile strength.

• 한국자기학회지
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• 제8권5호
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• pp.249-254
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• 1998

자기측정법에 의한 조사손상 평가 가능성을 조사하기 위하여 SA508 CI.3강 모재금속의 중성자 조사 및 열처리 온도에 따른 자기특성(자기이력곡선, Barkhausen Noise(BN) 진폭, BN 에너지)과 경도 변화를 측정 비교하였다. 중성자 조사에 따라 자화율, BN 진폭, BN 에너지는 감소하였고 보자력과 경도는 증가하였으며, 포화자화 값은 변화하지 않았다. 열처리된 조사시편의 경우, 열처리 온도 증가에 따라 BN 에너지는 증가하였으며, 경도는 감소하였다. 결함과 전위혹은 자벽 이동과의 상호작용에 의한 경도 및 자기특성의 일관성 있는 변화는 원자로 압력용기 재료의 조사손상 평가와 관련 자기적 측정법 응용 가능성을 보여주었다.

• Nuclear Engineering and Technology
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• 제24권2호
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• pp.163-167
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• 1992

고순도알루미늄중 불순물의 Parameter로 이용될수 있는 구리의 비파괴 방사화분석법의 고찰 및 23종의 극미량불순성분원소의 함량을 분석하였다. 즉 구리의 분석은 원자로의 속중성자에 의한 27Al(n,$\alpha$)24Na반응으로 생성되는 24Na의 방사능을 감소시키기 위하여 Thermal Column을 이용하였고 다른 조사공을 이용한 경우보다 약 100 배 정 도 방해 요인을 감소시킬 수 있었다. 24Na 에 의한 영향은 2-3 %범위 이하이었다. 이 방법에 의해 표준알루미늄(6 nine class)시료로부터 구리를 정량하였고 아울러 기타 불순원소들을 일상 방사화분석법에 의해 정량하였다. 구리의 함량은 0.54$\pm$0.08 ppm이었다. 이러한 결과는 문헌값과 비교할때 타당성이 있었고 일상분석에 이용할 수 있는 좋은 방법으로 여겨진다.

• Nuclear Engineering and Technology
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• 제30권4호
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• pp.364-376
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• 1998

The statistical analysis method was applied to the evaluation of fracture toughness in the ductile-brittle transition temperature region. Because cleavage fracture in steel is of a statistical nature, fracture toughness data or values show a similar statistical trend. Using the three-parameter Weibull distribution, a fracture toughness vs. temperature curve (K-curve) was directly generated from a set of fracture toughness data at a selected temperature. Charpy V-notch impact energy was also used to obtain the K-curve by a $K_{IC}$ -CVN (Charpy V-notch energy) correlation. Furthermore, this method was applied to evaluate the neutron irradiation embrittlement of reactor pressure vessel (RPV) steel. Most of the fracture toughness data were within the 95% confidence limits. The prediction of a transition temperature shift by statistical analysis was compared with that from the experimental data.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.301-308
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• 2011

The diametral creep of pressure tubes (PTs) in CANDU (CANada Deuterium Uranium) reactors is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of the heat transport system (HTS). PT diametral creep leads to diametral expansion, which affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux (CHF). The CHF is a major parameter determining the critical channel power (CCP), which is used in the trip setpoint calculations of regional overpower protection (ROP) systems. Therefore, it is essential to predict PT diametral creep in CANDU reactors. PT diametral creep is caused mainly by fast neutron irradiation, temperature and applied stress. The objective of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. The linear model was optimized using a genetic algorithm and was devised based on a bundle position because it is expected that each bundle position in a PT channel has inherent characteristics. The proposed BPLM for predicting PT diametral creep was confirmed using the operating data of the Wolsung nuclear power plant in Korea. The linear model was able to predict PT diametral creep accurately.