• Nuclear Engineering and Technology
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• 제50권2호
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.

• Nuclear Engineering and Technology
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• 제40권2호
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• pp.117-126
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• 2008

In this paper, a new reactor core design is proposed on the basis of a mixed core concept consisting of a thermal zone and a fast zone. The geometric structure of the fuel assembly of the thermal zone is similar to that of a conventional thermal supercritical water-cooled reactor(SCWR) core with two fuel pin rows between the moderator channels. In spite of the counter-current flow mode, the co-current flow mode is used to simplify the design of the reactor core and the fuel assembly. The water temperature at the exit of the thermal zone is much lower than the water temperature at the outlet of the pressure vessel. This lower temperature reduces the maximum cladding temperature of the thermal zone. Furthermore, due to the high velocity of the fast zone, a wider lattice can be used in the fuel assembly and the nonuniformity of the local heat transfer can be minimized. This mixed core, which combines the merits of some existing thermal SCWR cores and fast SCWR cores, is proposed for further detailed analysis.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

• 한국생산제조학회지
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• 제22권3_1spc호
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• pp.573-579
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• 2013

Once-through helically coiled steam generator tubes subjected to external pressure are of interest because of their application to advanced small- and medium-sized integral reactors, in which a primary coolant with a relatively higher pressure flows outside the tubes, while secondary water with a relatively lower pressure flows inside the tubes. Another notable point is that the values of the mean radius to thickness ratio of these steam generator tubes are very small, which means that a thick-walled cylinder is employed for these steam generator tubes. In the present paper, the maximum allowable pressure of helically coiled and thick-walled steam generator tubes with through-wall cracks under external pressure is investigated based on a detailed nonlinear three-dimensional finite element analysis. In terms of the crack orientation, either circumferential or axial through-wall cracks are considered. In particular, in order to quantify the effect of the crack location on the maximum external pressure, these cracks are assumed to be located in the intrados, extrados, and flank of helically coiled cylinders. Moreover, an evaluation is also made of how the maximum external pressure is affected by the ovality, which might be inherently induced during the tube coiling process used to fabricate the helically coiled steam generator tubes.

• 한국진공학회지
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• 제6권S1호
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• pp.96-109
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• 1997

In order to simulate the corrosion behavior of Zr-4 alloy in pressurized water reactors it was implanted (or bombarded) with 190ke V $Zr^+\; and \;Ar^+$ ions at liquid nitrogen temperature and room temperature respectively up to a dose of $5times10^{15} \sim 8\times10^{16} \textrm{ions/cm}^2$ The oxidation behavior and electrochemical vehavior were studied on implanted and unimplanted samples. The oxidation kinetics of the experimental samples were measured in pure oxygen at 923K and 133.3Pa. The corrosion parameters were measured by anodic polarization methods using a princeton Applied Research Model 350 corrosion measurement system. Auger Electron Spectroscopy (AES) and X-ray Photoelectric Spectroscopy (XPS) were employed to investigate the distribution and the ion valence of oxygen and zirconium ions inside the oxide films before and after implantation. it was found tat: 1) the $Zr^+$ ion implantation (or bombardment) enhanced the oxidation of Zircaloy-4 and resulted in that the oxidation weight gain of the samples at a dose of $8times10^{16}\textrm{ions/cm}^2$ was 4 times greater than that of the unimplantation ones;2) the valence of zirconium ion in the oxide films was classified as $Zr^0,Zr^+,Zr^{2+},Zr^{3+}\; and \;Zr^{4+}$ and the higher vlence of zirconium ion increased after the bombardment ; 3) the anodic passivation current density is about 2 ~ 3 times that of the unimplanted samples; 4) the implantation damage function of the effect of ion implantation on corrosion resistance of Zr-4 alloy was established.

• 유기물자원화
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• 제19권1호
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• pp.64-78
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• 2011

본 연구에서는 염분이 높은 단무지폐수를 대상으로 새로운 침지형 막분리 장치를 제작하고 기존의 소수성이 강한 폴리에틸렌 재질의 비개질 분리막과 이 소수성 분리막에 이온빔을 조사하여 친수성을 높여준 표면개질 분리막을 사용한 성능실험에서 시간에 따른 플럭스(flux)와 압력 변화, 유기물과 부유물질, 영양염류 등 오염물질의 제거 특성을 알아보았다. 간헐식 비개질막을 사용한 실험 결과, 합성폐수와는 달리 실제폐수에서는 투과 압력이 급격히 증가하여 심한 파울링(fouling) 현상이 일어남을 알 수 있으며 이는 실제폐수에 존재하는 첨가제 등 고분자물질과 염분에 의한 영향으로 추정된다. 약품세정 후의 막과 물세정 후의 막의 압력과 플럭스 변화 실험에서 오염된 막을 재생하기 위해 약품세정이 반드시 필요하며 막 운전시 연속식보다는 간헐식으로 운전하는 것이 성능이 더 우수함을 알 수 있었다. 또한 개질막의 경우가 비개질막의 경우보다 파울링에 도달하는 시간이 약 6배가 크므로 개질막의 경우가 막의 교체 비용을 1/6로 줄일 수 있다. 표면개질 분리막과 비개질 분리막 모두 처리수는 대체로 양호한 수질을 나타내고 있으며 특히 SS 제거, 질소 및 인 제거에도 탁월한 성능을 나타내고 있다.

• Journal of Radiation Protection and Research
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• 제35권2호
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• pp.57-62
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• 2010

원전 계획예방정비기간 증기발생기 수실작업, 가압기 전열관교체 또는 압력관피더 제거작업 지역 등은 높은 방사선량률을 보이는 지역으로, 짧은 시간 동안의 작업으로 작업종사자는 높은 피폭을 받을 가능성이 있다. 특히, 방사성물질과 접촉하는 손 부위는 높은 피폭이 일어날 수 있다. 이런 점을 고려하여 국내 가압경수로원전과 가압중수로원전의 계획 예방정비기간 중 증기발생기 수실 노즐댐 설치와 제거작업, 원자로 냉각재펌프 보수작업, 원자로헤드 보수 및 검사작업 등과 같은 고피폭 접촉작업에서 방사선작업종사자의 말단선량을 측정하기위한 현장시험을 실시하였다. 여기에 참여한 작업종사자는 가슴과 등 부위에 일상적인 절차에 따른 복수선량계를 패용한 것 이외에 손목에 개인선량계를 추가로 패용하였고, 손가락 부위에는 말단선량계 (Extremity dosimeter)를 패용하였다. 그 결과, 손가락이 받는 등가선량은 각각 손목이 받는 등가선량 및 가슴 또는 등 부위가 받는 등가선량과 일정한 비율로 평가됨을 확인하였다.

• 한국전산유체공학회지
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• 제13권4호
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• pp.86-95
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• 2008

For analyses of multi-phase flows in a water-cooled nuclear power plant, a three-dimensional SIMPLE-algorithm based hydrodynamic solver CUPID-S has been developed. As governing equations, it adopts a two-fluid three-field model for the two-phase flows. The three fields represent a continuous liquid, a dispersed droplets, and a vapour field. The governing equations are discretized by a finite volume method on an unstructured grid to handle the geometrical complexity of the nuclear reactors. The phasic momentum equations are coupled and solved with a sparse block Gauss-Seidel matrix solver to increase a numerical stability. The pressure correction equation derived by summing the phasic volume fraction equations is applied on the unstructured mesh in the context of a cell-centered co-located scheme. This paper presents the numerical method and the preliminary results of the calculations.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.530-537
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• 2020

This study investigates the effect of an included angle and heat flux on heat transfer of V-shape tube array having a horizontal upper tube. The test uses two stainless steel tubes with a smooth surface submerged under the water at atmospheric pressure. The angle varies from 2° to 24°. The heat transfer coefficient gets decreasing in consequence as the angle increases. The enhancement due to the lower tube is distinct as the heat flux is lower than 60 kW/㎡, where the effect of the convective flow is dominant. The present study and the published results show a similar tendency. Although the heat transfer coefficient for the present study is smaller than the symmetry case, enhanced heat transfer is observed compared to the tube array having a lower horizontal tube as the included angle is less than 10°.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 학술대회
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• pp.71-78
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• 2008

For analyses of multi-phase flows in a water-cooled nuclear power plant, a three-dimensional SIMPLE-algorithm based hydrodynamic solver CUPID-S has been developed. As governing equations, it adopts a two-fluid three-field model for the two-phase flows. The three fields represent a continuous liquid, a dispersed droplets, and a vapour field. The governing equations are discretized by a finite volume method on an unstructured grid to handle the geometrical complexity of the nuclear reactors. The phasic momentum equations are coupled and solved with a sparse block Gauss-Seidel matrix solver to increase a numerical stability. The pressure correction equation derived by summing the phasic volume fraction equations is applied on the unstructured mesh in the context of a cell-centered co-located scheme. This paper presents the numerical method and the preliminary results of the calculations.