• Nuclear Engineering and Technology
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• 제27권5호
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• pp.775-783
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• 1995

CANDU원자로에서 압력관의 건전성을 향상시키기 위한 방안으로 압력관의 두께를 증가시키는 방법과 압력관 제조공정에서 초기수소농도를 줄이는 방법이 연구중에 있다. 본 연구에서는 압력관 두께증가가 가동중 압력관의 안전여유도에 미치는 영향과 새로운 압력관의 낮은 수소농도가 파손의 주원인인 DHC에 미치는 영향에 대해 연구하였다. 가동중 압력관에 날카로운 결함이 발생할 경우 5.2mm두꺼운 압력관은 안전여유도 관점에서 현행 2mm 압력 관에 비해 25% 증가효과를 보이는 것으로 나타났다. LBB평가에서도 두꺼운 압력관은 DHC 발생에 필요한 초기균열 길이 (a), 중수누설 감지 시점에서의 균열길이 (Lp), 중수누설후 압력관 파단까지의 허용시간(t)등에서 많은 이점이 있는 것으로 평가되었으며, 또한 LOCA시 압력관 파단관점에서도 유익한 것으로 나타났다. 여러가지 다른 두께 및 다른 초기수소농도를 갖는 압력관을 대상으로 20년 가동후의 총 누적 수소량을 계산한 결과, 5ppm의 초기 수소량을 갖고 두께가 5.2 mm인 압력관이 가장 우수한 저항성을 보였다. 결함 성장평가에 있어서 초기에 낮은 수소량을 갖는 압력관은 20년 가동후에도 수소화물의 석출이 일어나는 TSS 도달 온도가 낮게 유지되며 냉각시 균열성장량도 매우 적은 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제22권4호
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• pp.359-370
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• 1990

이 논문은 원자력발전소의 안전사고해석에 흔히 이용되는 중성자 다군확산 동특성방정식의 저차원(0차원 및 1차원) 수치해를 3차원 수치해와 비교함으로써 저차원 수치해법에 요구되는 동특성해석 입력자료를 체계적으로 유도하기 위한 것이다. 이 목적으로 이 논문에서는 수정형 Borresen 소격모형에 의한 3차원 동특성 해석코드인 CMSNACK 전산코드로 LRA-BWR 경수로 동특성 시범문제의 3차원해를 구하고 이 해를 기준으로 삼아 중성자 다군확산 동특성방정식의 1차원 유한차분해와 3차 Hermit 다항식 전개해법에 의한 점운동방정식의 0차원 수치해를 비교하고자 했다. 중성자 다군확산방정식의 1차원 유한차분해와 점운동방정식의 0차원 수치해를 구하기 위해 ODTRAN 전산코드와 POTRAN 전산코드를 개발하였고 이들 코드의 입력자료는 ODTRAN 코드의 경우 중성자속 체적가중법을 POTRAN의 경우 단열근사법을 수정하여 마련하였다. 이같이 마련한 입력자료를 써서 LRA-BWR 동특성문제에 대한 1차원 및 0차원 해를 구했으며 그 결과를 CMSHACK코드에 의한 3차원 해와의 비교를 통해서 저차원 수치해의 계산효율성과 안전해석코드에 요구되는 계산결과의 보수성 등을 조사했다. 이같은 비교결과를 토대로 저차원 수치해법의 입력자료 마련에 이 논문에서 제시한 방법이 유용하게 이용될 수 있음을 보였다.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.19-24
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• 2011

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, $1.1{\times}10^{21}\;n/cm^2$). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1162-1173
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• 2016

In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium-water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional thermally nonequilibrium three-component gas-liquid flow model (sodium-hydrogen-sodium hydroxide). Consideration is given to the results of a mathematical modeling of experiments involving steam injection into the sodium loop of a circulation test facility. That was done by means of the computer code in which the proposed model had been implemented.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1964-1969
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• 2019

The safety and efficiency of nuclear reactors largely depend on the monitoring and control of nuclear radiation. Due to the unique nuclear-physical characteristics, Hf is one of the most promising materials for the manufacturing of the control rods and the emitters of neutron detectors. It is proposed to use the Compton neutron detector with the emitter made of Hf in the In-core Instrumentation System (ICIS) for monitoring the neutron field. The main advantages of such a detector in comparison the conventional β-emission sensors are the possibility of reaching of a higher cumulative radiation dose and the absence of signal delays. The response time of the detection is extremely important when a nuclear reactor is operating near its critical operational parameters. Taking Hf as an example, the general principles for calculating the chains of materials transformation under neutron irradiation are reported. The influence of ^179m1Hf on the Hf composition changing dynamics and the process of transmutants' (Ta, W) generation were determined. The effect of these processes on the absorbing properties of Hf, which inevitably predetermine the lifetime of the detector and its ability to generate a signal, is estimated.

• 태양에너지
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• 제18권3호
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• pp.197-203
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• 1998

A 5 ton/day pilot scale two-phase anaerobic digester was constructed and tasted to treat Korean food wastes in Anyang city. The process was developed based on 3 years of lab-scale experimental results on am optimim treatment method for the recovery of biogas and humus. Problems related to food waste are ever Increasing quantity among municipal solid wastes(MSW) and high moisture and salt contents. Thus our food waste produces large amounts of leachate and bed odor in landfill sites which are being exhausted. The easily degradable presorted food waste was efficiently treated in the two-phase anaerobic digestion process. The waste contained in plastic bags was shredded and then screened for the removal of inert material such as fabrics and plastics, and subsequently put into the two-stage reactors. Heavy and light inerts such as bones, shells, spoons and plastic pieces were again removed by gravity differences. The residual organic component was effectively hydrolyzed and acidified in the first reactor with 5 days space time at pH of about 6.5. The second, methanization reactor part of which is filled with anaerobic fillters, converted the acids into methane with pH between 7.4 to 7.8. The space time for the second reactor was 15 days. The effluent from the second reactor was recycled to the first reactor to provide alkalinities. The process showed stable steady state operation with the maximum organic rate of 7.9 $kgVS/m^3day$ and the volatile solid reduction efficiency of about 70%. The total of 3.6 tons presorted MSW containing 2.9 tons of food organic was treated to produce about $230m^3$ of biogas with 70% of methane and 80kg humus. This process is extended to full scale treating 15 tons of food waste a day in Euiwang city and the produced biogas is utilized for the heating/cooling of adjacent buildings.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.1009-1014
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• 2016

SiC-based ceramic composites are currently being considered for use in fuel cladding tubes in light-water reactors. The joining of SiC ceramics in a hermetic seal is required for the development of ceramic-based fuel cladding tubes. In this study, SiC monoliths were diffusion bonded using a Ti foil interlayer and additional Si powder. In the joining process, a very low uniaxial pressure of ~0.1 MPa was applied, so the process is applicable for joining thin-walled long tubes. The joining strength depended strongly on the type of SiC material. Reaction-bonded SiC (RB-SiC) showed a higher joining strength than sintered SiC because the diffusion reaction of Si was promoted in the former. The joining strength of sintered SiC was increased by the addition of Si at the Ti interlayer to play the role of the free Si in RB-SiC. The maximum joint strength obtained under torsional stress was ~100 MPa. The joint interface consisted of $TiSi_2$, $Ti_3SiC_2$, and SiC phases formed by a diffusion reaction of Ti and Si.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.427-446
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• 2009

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical $CO_2$ Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.