• 분석과학
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• 제12권6호
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• pp.490-497
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• 1999

불활성기체용융-열전도도 측정법의 수소분석기를 개조하여 글로브박스에 설치했으며 조사핵연료피복관의 수소분석에 사용했다. Zr과 Ti 매질의 수소표준물질로 용융조제인 주석을 사용하지 않고도 수소함량 $3{\mu}g$까지 분석가능하였다. 시료의 크기가 작을수록 수소 회수율이 높았으며 지르칼로이 시료의 수소분석시 Ti 매질의 표준물질을 사용할 수 있음을 확인하였다. 수소분석에 사용한 실제 조사핵연료피복관의 평균 방사능은 10 mR/hr였으며 평균수소농도는 130 ppm이었다.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1210-1217
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• 2021

The analysis of ⁴¹Ca in concrete generated from the nuclear facilities decommissioning is critical for ensuring the safe management of radioactive waste. An analytical method for the determination of ⁴¹Ca in concrete is described. ⁴¹Ca is a neutron-activated long radionuclide, and hence, for accurate analysis, it is necessary to completely extract Ca from the concrete sample where it exists as the predominant element. The decomposition methods employed were the acid leaching, microwave digestion, and alkali fusion. A comparison of the results indicated that the alkali fusion is the most suitable way for the separation of Ca from the concrete sample. Several processes of hydroxide and carbonate precipitation were employed to separate ⁴¹Ca from interferences. The method relies on the differences in the solubility of the generated products. The behavior of Ca and the interfering elements such as Fe, Ni, Co, Eu, Ba, and Sr is examined at each separation step. The purified ⁴¹Ca was measured by a liquid scintillation counter, and the quench curve and counting efficiency were determined by using a certified reference material of known ⁴¹Ca activity. The recoveries in this study ranged from 56 to 68%, and the minimum detectable activity was 50 mBq g^-1 with 0.5 g of concrete sample.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.791-796
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• 2020

Nuclear fusion seems to be a good choice of energy source in the future. Nickel is one of the crucial structural materials for fusion devices. In this work, the cross section data of ⁵⁸Ni(n,p)⁵⁸Co, ⁶⁰Ni(n,p)⁶⁰Co, ⁶¹Ni(n,p)⁶¹Co, ⁶²Ni(n,p)⁶²Co and ⁶⁴Ni(n,p)⁶⁴Co reactions were calculated using the nuclear codes ALICE/ASH, EMPIRE 3.2 and TALYS 1.8. In addition, the cross sections were calculated with the empirical formulas obtained in our previous paper at 14-15 MeV. The obtained results were compared with the measured values in the literature, and with the evaluated data files (JEFF-3.3, TENDL-2017, ENDF/B-VIII.0).

• Nuclear Engineering and Technology
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• 제44권2호
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• pp.113-150
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• 2012

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

• Nuclear Engineering and Technology
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• 제41권1호
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• pp.1-10
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• 2009

The safe and reliable performance of fusion and fission plants depends on the choice of suitable materials and an assessment of long-term materials degradation. These materials are degraded by their exposure to extreme conditions; it is necessary, therefore, to address the issue of long-term damage evolution of materials under service exposure in advanced plants. The empirical approach to the study of structural materials and fuels is reaching its limit when used to define and extrapolate new materials, new environments, or new operating conditions due to a lack of knowledge of the basic principles and mechanisms present. Materials designed for future Gen IV systems require significant innovation for the new environments that the materials will be exposed to. Thus, it is a challenge to understand the materials more precisely and to go far beyond the current empirical design methodology. Breakthrough technology is being achieved with the incorporation in design codes of a fundamental understanding of the properties of materials. This paper discusses the multi-scale, multi-code computations and multi-dimensional modelling undertaken to understand the mechanical properties of these materials. Such an approach is envisaged to probe beyond currently possible approaches to become a predictive tool in estimating the mechanical properties and lifetimes of materials.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2728-2745
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• 2021

The purpose of this study is to reduce seismic responses of an actual nuclear piping system using a tuned mass damper (TMD) device. A numerical piping model was developed and validated based on shaking table test results with actual nuclear piping. A TMD for nuclear piping was newly devised in this work. A TMD shape design suitable for nuclear piping systems was conducted, and its operating performance was verified after manufacturing. The response reduction performance of the developed TMD under earthquake loading on actual piping was investigated. Results confirmed that, on average, seismic response reduction rates of 34% in the maximum acceleration response, 41% in the root mean square acceleration response, and 57% in the spectral acceleration response were shown through the TMD application. This developed TMD operated successfully within the seismic response reduction rate of existing TMD optimum design values. Therefore, the developed TMD and dynamic interpretation help improve the nuclear piping's seismic performance.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.1962-1971
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• 2022

The liquid lead-lithium (Pb-17Li) blanket has many applications in fusion reactors due to its good tritium breeding performance, high heat transfer efficiency and safety. The compatibility of liquid Pb-17Li alloy with the structural material of blanket under magnetic field is one of the concerns. In this study, corrosion experiments China low activation martensitic (CLAM) steel and 316L steel were carried out in a forced convection Pb-17Li loop under 1.0 T magnetic field at 480 ℃ for 1000 h. The corrosion results on 316L steel showed the characteristic with a superficial porous layer resulted from selective leaching of high-soluble alloy elements and subsequent phase transformation from austenitic matrix to ferritic phase. Then the porous layers were eroded by high-velocity jet fluid. The main corrosion mechanism of CLAM steel was selective dissolution-base corrosion attack on the microstructure boundary regions and exclusively on high residual stress areas. CLAM steel performed a better corrosion resistance than that of 316L steel. The high Ni dissolution rate and the erosion of corroded layers are the main causes for the severe corrosion of 316L steel.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2678-2686
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• 2023

This work demonstrates the results of full-scale numerical experiments of a hybrid thorium-containing fuel plant operating in a state close to critical due to a controlled source of D-T neutrons. The proposed facility represented a level of generated power (~10-100 MW_t) in a small pilot. In this work, the simulation of the D-T neutron plasma source operation in conjunction with the facility blanket was performed. The fission of fuel nuclei and the formation of spatial-energy release were studied in this simulation, in pulsed and stationary modes of the facility operation. The optimization results of neutronic and fluid dynamics studies to level the emerging offsets of the radial energy formed in the volume of the facility multiplying part due to the pulsed operation of the D-T neutron plasma source were presented. The results will be useful in improving the power control-based subcriticality monitoring method in coupled systems of the "pulsed neutron source-subcritical fuel assembly" type.

• 한국진공학회지
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• 제14권2호
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• pp.69-77
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• 2005

중성입자빔 가열장치의 이온원과 빔라인 부품들을 개발하고 시험하기 위해 제작한 부피 $60m^3$의 시험설비에는 특성 비교를 위해 몇 가지 다른 방식으로 제작된 대용량 크라이오 흡착펌프를 장착하여 사용하고 있다. 크라이오 펌프의 활성탄 패널을 냉각시키면서 수소를 적절한 간격으로 도입하여 수소 분압을 측정하고 그 시간적인 변화를 통해 온도에 따른 수소흡착과 방출특성을 분석해 보았으며 관련 파라미터 사이의 상호 영향에 대해 알아보기 위해 시뮬레이션을 수행했다.

• 한국수소및신에너지학회논문집
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• 제23권1호
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• pp.49-55
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• 2012

Hydrogen isotopes accountancy and storage are important functions in a nuclear fusion fuel cycle. The hydrogen isotopes are safely stored in metal hydride beds. The tritium inventory of the bed is determined from the decay heat of tritium. The decay heat is measured by circulating helium through the metal hydride bed and measuring the resultant temperature increase of the helium flow. We are reporting our preliminary experimental results on the hydrogen isotopes accountancy and storage performance in a metal hydride bed.