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하나로에서의 NTD 조사를 위한 중성자속 평탄화 장치의 최적화를 위한 예비분석
송영동,이헌주,이병택,전병진,김학노 濟州大學校 産業技術硏究所 2001 산업기술연구소논문집 Vol.12 No.1
NTD (Pieutron Transmutation Doping) method has several advantages of high resistivity and uniform doping in comparison with other method. To satisfy those conditions. the flux variations of radial and axial directions should be within ±5 % and ±1.7%. respectively. The NTD facility in HANARO is purposed to irradiate the silicon ingot of 60cm in height. Hence. the flux flattener will be designed for 60cm. In this paper. preliminary study for optimal design is showed and the flux distribution of axial direction is calculated using MCNP4B code. The results show that the flattener model can flatten the flux to 83% of total length.
하나로에서의 NTD조사를 위한 중성자속 평탄화 장치의 최적화를 위한 예비분석
송영동,이헌주,이병철,전병진,김학노 제주대학교 산업기술연구소 2001 尖端技術硏究所論文集 Vol.12 No.1
NTD(Neutron Transmutation Doping) method has several advantages of high resistivity and uniform doping in comparison with other method. To satisfy those conditions, the flux variations of radial and axial directions should be within ±5% and ±1.7%. respectively. The NTD facility in HANARO is purposed to irradiate the silicon ingot of 60cm in height. Hence, the flux flattener will be designed for 60cm. In this paper, preliminary study for optimal design is showed and the flux distribution of axial direction is calculated using MCNP4B code. The results show that the flattener model can flatten the flux to 83% of total length.
Utilization of the Stand-by Fuel Assemblies
Kim, Hark-Rho,Chung, Chang-Hyun Korean Nuclear Society 1981 Nuclear Engineering and Technology Vol.13 No.2
The change in the design-basis refueling strategy caused by the unexpected nuclear fuel failures may result in discharging intact fuel assemblies which were irradiated in the positions symmetric to the failed ones in addition to the failed ones in order to maintain the symmetric power shape in the reactor core. In this work an attempt is made to reuse the intact fuel assemblies which were discharged before reaching the design turnup in the above-described situation so as to improve the fuel utilization. The TDCORE code is used to estimate the flux and power distribution, and the RELOAD-II code for searching the optimal loading pattern with the minimum assembly radial power peaking factor. For the case of the Ko-ri unit 1, its third cycle turnup could be extended to 11,648 MWD/MTU by reusing the four low-burned fuel assemblies removed at the end of the first cycle, and then the loading pattern is searched to the equilibrium cycle.
Homogenization of KMRR Hafnium Control Assembly for 3-D Diffusion Calculation
Park, Hang-Bok,Kim, Young-Jin,Kim, Hark-Rho,Lee, Ji-Bok Korean Nuclear Society 1988 Nuclear Engineering and Technology Vol.20 No.4
The hafnium shroud is used to control the excess reactivity and power distribution in KMRR. The core analysis is performed by the diffusion code VENTURE using the 5 group macroscopic cross sections homogenized for an assembly. Investigated are the applicability of the diffusion calculation by homogenized cross sections to the analysis of control assembly which features unusual geometry such that hafnium shroud surrounds a multiplying medium inside. Comparative calculation is performed for the excess reactivity and power levels by the transport code TWOTRAN. The results show the acceptability of the diffusion calculation by the homogenized cross sections without significant error.
OPPORTUNITIES AND CHALLENGES OF NEUTRON SCIENCE AND TECHNOLOGY IN KOREA
KYE HONG LEE,J. M. SUNGIL PARK,HARK-RHO KIM,BYUNG JIN JUN,YOUNG-JIN KIM,JAE-JOO HA,김만원,최성민 한국원자력학회 2009 Nuclear Engineering and Technology Vol.41 No.4
Neutron science and technology, the utilization of neutron beams for a wide variety of scientific and engineering research ranging from materials and life science to industrial applications, has been one of the key elements of modern science and technology. Currently, the neutron science and technology in Korea is in rapid growth with the operation of the 30 MW High-flux Advanced Neutron Application Reactor (HANARO) at the Korea Atomic Energy Research Institute, which is one of the most powerful nuclear research reactors in the world. Furthermore, a state of the art HANARO cold neutron research facility, which will open a new era for the neutron science and technology in Korea, is expected to become available in 2010. In this paper, the progress of neutron science and technology in Korea is reviewed and its unprecedented new opportunities and challenges in coming years are presented.
Chung, Yong-Sam,Kim, Sun-Ha,Moon, Jong-Hwa,Kim, Hark-Rho,Kim, Young-Jin Korean Nuclear Society 2006 Nuclear Engineering and Technology Vol.38 No.6
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.
Pin Power Reconstruction of HANARO Fuel Assembly via Gamma Scanning and Tomography Method
Seo, Chul-Gyo,Park, Chang-Je,Cho, Nam-Zin,Kim, Hark-Rho Korean Nuclear Society 2001 Nuclear Engineering and Technology Vol.33 No.1
To determine the pin power distribution without disassembling, HANARO fuel assemblies are gamma-scanned and then the distribution is reconstructed tv using the tomography method. The iterative least squares method (ILSM and the wavelet singular value decomposition method (WSVD) are chosen to solve the problem. An optimal convergence criterion is used to stop the iteration algorithm to overcome the potential divergence in ILSM. WSVD gives better results than ILSM , and the average values from the two methods give the best results. The RMSE (root mean square errors) to the reference data are 5.1, 6.6, 5.0, 6.5, and 6.4% and the maximum relative errors are 10.2, 13.7, 12.2, 13.6, and 14.3%, respectively. It is found that the effect of random positions of the pins is important. Although the effect can be accommodated by the iterative calculations simulating the random positions, the use of experimental equipment with a slit covering the whole range of the assembly horizontally is recommended to obtain more accurate results. We made a new apparatus using the results of this study and are conducting an experiment in order to obtain more accurate results.
Generation and Benchmarking of a 69-group Cross Section Library for Thermal Reactor Applications
Kim, Jung-Do,Lee, Jong-Tai,Gil, Choong-Sup,Kim, Hark-Rho Korean Nuclear Society 1989 Nuclear Engineering and Technology Vol.21 No.4
A 69-group cross section library consisting of more than 130 materials was generated for thermal reactor applications using the NJOY nuclear data processing system and the recent version of evaluated nuclear data files available from IAEA Nuclear Data Section. The multigroup library was validated through the analysis of various criticality experiments and depletion results of PWR. When used with the WIMS-KAERI code, the average $K_{eff}$ obtained for 47 uranium-oxide and 41 uranium metal fueled critical configurations is 0.9997 with a standard deviation of 0.69 percent. The calculated burnup dependent isotopic inventories of uranium and plutonium generally show good agreement with measured values obtained from depleted PWR pins.s.