• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1251-1260
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• 2019

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to $7MW/m^2$ while the maximum operated temperature of Eurofer97 is set to $550^{\circ}C$. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to $0.75MW/m^2$. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to $1.69MW/m^2$ only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

• 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.

• Korean Chemical Engineering Research
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• 제50권4호
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• pp.595-603
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• 2012

국제핵융합실험로(ITER)가 2019년까지 7개국의 공동개발사업으로 건설될 예정이다. ITER의 핵융합연료주기는 핵융합진공용기, 삼중수소 플랜트, 연료공급부로 구성되어 있다. 이중에서 삼중수소 플랜트는 핵융합연료주기를 위한 중 수소와 삼중수소의 저장, 공급, 분리, 제거, 회수 등의 기능을 제공한다. 삼중수소 플랜트는 외부에서 중수소와 삼중수소를 공급받아 저장 공급하는 SDS, 토카막배출처리의 TEP, 수소동위원소 분리의 ISS, 삼중수소수 및 대기 처리의 WDS ADS, 정성 정량분석의 ANS 등으로 구성된다. 이 논문에서는 삼중수소 플랜트를 구성하는 주요 공정에 대한 기능 및 설계요건을 기술하였다. 한국은 SDS 개발에 참여하고 있으며 월성원전 삼중수소 제거설비(WTRF) 건설 및 운전경험을 통해 WDS 대한 기술을 일부 확보하였다. 향후 ISS 및 TEP에 대한 기술확보를 위한 여러 분야에서의 참여 확대를 기대하고 있다.

• Journal of Welding and Joining
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• 제14권5호
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• pp.122-133
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• 1996

In order to evaluate the cryogenic fracture characteristics of structural steels for superconducting magnets of fusion reactor, small punch (SP) testing was performed on austenitic stainless steel (JN1 base metal) and its TIG weldments at 293K, 77K and 4K. The mechanical properties with respect to the extracted location of the weld metal, on the effects of welding heat cycle about base metal near fusion line in TIG weldments were investigated. The mechanical property of the weld metal in TIG weldments depends on distance from welding root, root region of weldments having the lowest mechanical property. The base metal near fusion line showed degradation of mechanical property caused by cyclic heating during the TIG welding. Based on the test results, HAZ was found to be up to 5mm from the fusion line. It is shown that SP testing is a useful tool to evaluate the mechanical properties with respect to the microstructures changes such as HAZ as well as weld metal in TIG weldments at cryogenic temperature.

• Journal of Welding and Joining
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• 제24권2호
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• pp.48-56
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• 2006

The fatigue strength of the welds is affected by such factors as the weld geometry, microstructures, tensile properties and residual stresses caused by fabrication. It is very important to evaluate the structural integrity of the welds in nuclear power plant because the weldment undergoes the most of damage and failure mechanisms. In this study, the fatigue assessments for a reactor vessel outlet nozzle with the weldment to the piping system are performed considering the welding residual stresses as well as the effect of local brittle zone in the vicinity of the weld fusion line. The analytical approaches employed are the microstructure and mechanical properties prediction by semi-analytical method, the thermal and stress analysis including the welding residual stress analysis by finite element method, the fatigue life assessment by following the ASME Code rules. The calculated results of cumulative usage factors(CUF) are compared for cases of the elastic and elasto-plastic analysis, and with or without residual stress and local brittle zone effects, respectively. Finally, the fatigue life of reactor vessel outlet nozzle weld is slightly affected by the local brittle zone and welding residual stresses.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2004년도 Proceedings of the 4th Korea-China Joint Workshop on Nuclear Waste Management
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• pp.220-237
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• 2004

China, as a developing country with a great number of population and relatively less energy resources, reasonably emphasizes the nuclear energy utilization development. For the long term sustainable energy supply, as for nuclear application the basic strategy of PWR-FBR-Fusion has been settled and envisaged. Due to the economy and experience reasons the nuclear power and technology development with a moderate style are kept in China up to now. In China mainland apart from two NPPs with the total capacity of 2.1 GWe in operation, four NPPs are under construction and two NPPs are planned for the Tenth Five Year Plan(2001-2005). Also another one or two NPPs are still in discussion. It could be foreseen that the total nuclear power capacity will reach 8.5GWe before the year 2005 and 14-15 GWe before 2010 respectively. As the first step for the Chinese fast reactor engineering development the 65MWt China Experimental Fast Reactor(CEFR) is under construction. The main components of primary, secondary and tertiary circuits and of fuel handling system have been ordered. The reactor building under construction has reached the top namely 57m above the ground. More than one hundred components and shielding doors have been installed. It is planned that the construction of reactor building with about 40,000$m^2$ floor surface will be completed in the end of the year 2002 and envisaged that the first criticality of the CEFR will be in the end of 2005. The second step of the Chinese fast reactor engineering development is a 300MWe Prototype Fast Breeder Reactor which is only under consideration up to now. Some important technical selections have been settled, but its design has not yet started.

• Journal of Welding and Joining
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• 제21권1호
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• pp.87-92
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• 2003

JLF-1 steel (Fe-9Cr-2W-V-Ta), reduced activation ferritic steel, is one of the promising candidate materials for fusion reactor applications. Tensile properties of JLF-1 base metal and its TIG weldments has been investigated at the room temperature, $400^{\circ}C$ and $600^{\circ}C$. The tensile strength of base metal (JLF-1) showed the level between those of weld metal and the Heat Affected Zone (HAZ). When the test temperature was increased from room temperature to high temperature ($400^{\circ}C$ and $600^{\circ}C$), both strength and ductility decreased or base metal, weld metal and the HAZ. The longitudinal specimens of base metal represented similar strength and ductility at room temperature and high temperature, compared to those of transverse specimens. Little anisotropy for the rolling direction was observed in the base metal of JLF-1 steel.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.867-868
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• 2004

The maximum width of HAZ of SA508치.3 steel produced by overlay RPV cladding was approximately 10 mm and it was composed of variety of microstructures with various grain size and precipitates. In addition, along the weld fusion line there formed a heavy carbide precipitation zone in the width of $20{\sim}30\;{\mu}m$. 2. As the specimen sampling position approached to the weld fusion line, the increase in yield and tensile strength was approximately 90 and 40 MPa, respectively. Meanwhile, the plastic fracture strain reduced from 14 to 8 percent. 3. The lowest SP energy and the highest ductile to brittle transition temperature in the HAZ were observed at the coarse- and fine-grained HAZ.

• 한국지능시스템학회논문지
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• 제5권3호
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• pp.73-82
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• 1995

본 논문에서는 동특성식의 비선형이며 불확실성을 갖는 Tokamak 핵융합로의 온도와 밀도 제어를 위한 적응 퍼지제어 알고리즘을 개발하였다. Tokamak 핵융합로 동특성식의 불확실성을 매개변수적이 아니고 상태의존적이다. 따라서 기존의 비선형제어 방식으로는 다루기 힘든 어려움이 따른다. 제안된 적응 퍼지 제어기는 하나의 해결방법으로 사용될 수 있을 것이며 시뮬레이션을 통해 미리 지정된 운전영역 내에서는 만족할 만한 제어성능을 발휘함을 확인할 수 있었다.

• 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.