• Nuclear Engineering and Technology
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• 제2권2호
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• pp.85-95
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• 1970

두개의 threshold detector로서 인자로의 폭발사고시에 방출되는 속 중성자의 속도분포를 측정하고 그로부터 속 중성자의 인체흡수선량을 계산하였다. 이때 속 중성자의 속도분포는 하나의 스펙트럼 매개변수에 의하여 결정된다는 가정으로부터 얻어지는데 이 매개변수는 threshold detector의 반응율을 측정하므로서 구해진다. 속 중성자의 인체흡수선량은 속 중성자의 속도분포 변화에 따라 큰 변동이 없었으나 threshold detector의 평균반응단면적은 크게 변하였다. 따라서 속 중성자의 속도분포에 관계없이 threshold detector의 평균반응단면적을 고정된 값으로 취하여 속 중성자선량을 계산한다면 큰 오차를 일으키게 될 것이라는 것을 보여주었다. 한편 핵분열에서 방출되는 속 중성자의 속도분포에 대한 세 해석적 표현인 즉 Watt, Cranberg및 Maxwellian 공식들로부터 속 중성자 선량을 계산하여 서로 비교하였다. Watt 및 Cranberg 공식들로 부터 얻어진 속 중성자선량은 Maxwellian 공식으로부터 얻어진 그것보다 약간 높은 값을 보여 주었으며 Watt 공식에 의한 선량계산치는 Cranberg 공식에 의한 그것과 비슷한 값을 보여주었다.

• Nuclear Engineering and Technology
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• 제5권4호
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• pp.291-309
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• 1973

TRIGA Mark II와 III 원자로의 여러가지 가동조건에 있어서 노벽으로 부터의 누설 ${\gamma}$선에 의한 조사선양률을 3"$\times$3"원통형 NaI(T1) 섬광계수기와 400 channel파 고분석장치로 측정하였는데 측정된 spectrum으로부터 조사선양률을 산출하는데는 실제적면에서 복잡하기 짝이 없는 response matrix 방법대신 정도가 좋으면서도 비교적 그 과정이 단순한 Moriuchi의 specturm -조사선양률 환산 이론을 적용하였다. 연구결과에 따르면 노심에서 발생된 누설 ${\gamma}$선의 기본적인 spectrum 형태는 원자로의 열출력이나 차장벽에 의한 강도의 감쇠에 별로 영향을 받지 않고 있으며 원자로 누설${\gamma}$선에 의란 전조사선양률의 공기중에서의 감쇠는 폭 넓은 energy분포에도 불구하고 지수함수적 감쇠를 하고 있음이 판명되있다. 이 전조사선양률은 원자로의 열출력에 대체로 비례하고 있으나 TRIGA Mark III과 같은 가동형노심의 경우는 측정된 spectrum이 매우 다양한바, 그로부터 산출된 전조사선양률의 크기에는 관계없이, spectrum 분해방법을 적용하여 노심에서 발생된 누설 ${\gamma}$선과 원자로가동중 발생되는 여지 ${\gamma}$선의 기여를 판별 해석하는데 성공하였다.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.700-711
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• 2015

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called "self-wastage phenomena." A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.871-894
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• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.211-218
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• 2013

Large quantities of irradiated graphite waste from graphite-moderated nuclear reactors exist and are expected to increase in the case of High Temperature Reactor (HTR) deployment [1,2]. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 ($^{14}C$), with a half-life of 5730 years. Fachinger et al. [2] have demonstrated that thermal treatment of irradiated graphite removes a significant fraction of the $^{14}C$, which tends to be concentrated on the graphite surface. During thermal treatment, graphite surface carbon atoms interact with naturally adsorbed oxygen complexes to create $CO_x$ gases, i.e. "gasify" graphite. The effectiveness of this process is highly dependent on the availability of adsorbed oxygen compounds. The quantity and form of adsorbed oxygen complexes in pre- and post-irradiated graphite were studied using Time of Flight Secondary Ion Mass Spectrometry (ToF-SIMS) and Xray Photoelectron Spectroscopy (XPS) in an effort to better understand the gasification process and to apply that understanding to process optimization. Adsorbed oxygen fragments were detected on both irradiated and unirradiated graphite; however, carbon-oxygen bonds were identified only on the irradiated material. This difference is likely due to a large number of carbon active sites associated with the higher lattice disorder resulting from irradiation. Results of XPS analysis also indicated the potential bonding structures of the oxygen fragments removed during surface impingement. Ester- and carboxyl-like structures were predominant among the identified oxygen-containing fragments. The indicated structures are consistent with those characterized by Fanning and Vannice [3] and later incorporated into an oxidation kinetics model by El-Genk and Tournier [4]. Based on the predicted desorption mechanisms of carbon oxides from the identified compounds, it is expected that a majority of the graphite should gasify as carbon monoxide (CO) rather than carbon dioxide ($CO_2$). Therefore, to optimize the efficiency of thermal treatment the graphite should be heated to temperatures above the surface decomposition temperature increasing the evolution of CO [4].

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.1-5
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• 2017

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor ($K_{eff}$) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of $D_2O$, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

• 생명과학회지
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• 제10권1호
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• pp.14-21
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• 2000

음요수 중의 유기탄소의 제거수단으로써 biofiltration법의 타당성을 검토하였다.NOM에서의 생분해 가능한 분율을 알아보기위하여 행하여 졌다. UV 조사량을 3가지로 달리하였을 경우에 회분식에서의 생분해능과 비교하였다. 생물여과 반응기 실험의 경우에 생분해 특성을 검토한 결과 EBCT, 순환비 및 유입농도등과 같은 운전피라미터들이 생물여과 반응기의 생분해능에 영향을 미침을 알 수있었다. 생물여과 반응기의 유출수에서의 UV/DOC의 비는 반응기에 공급되는 원료중의 UV/DOC비에 비해 증가하였으며, 이로부터 생분해에 의해 저거된 DOC는 UV에 의해 그다지 흡수되지 않는 물질임을 알 수있었다. 본 여눅를 통해 생분해 가능한 DOC의 부분을 제거하는데 있어서 생물여과공법이 효과적인 방법이라는 것을 알수있었으며 UV처리와 bilfiltration을 연계한 공정은 수처리 시설에서 유출수의 DOC농도는 낮추는데 효과적인 한가지 방안으로 사료되었다.

• 멤브레인
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• 제2권1호
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• pp.21-32
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• 1992

화학증착법을 사용하여 다공성 Vycor 유리에 선택적 수소 투과성 실리카 막을 제조하였다. 화학증착에는 $SiCl_4$의 가수분해 반응이 이용되었으며, 반응물인 $SiCl_4$와 물을 서로 반대 방향으로 주입하여 막을 제조하는 opposing-reactants film deposition방법과, 반응물을 다공성 유리관의 한쪽으로만 공급하는 one-sided film deposition 방법을 모두 사용하였다. 제조된 실리카 막을 통한 수소의 투과도는 $600^{\circ}C$ 이상의 온도에서 $0.01-0.25cm^3(STP)/cm^2-min-atm$의 범위에 있었으며, 수소의 질소에 대한 투과도 비는 1000정도였고, 온도의 증가에 따라 실리카 막을 통한 수소 및 질소의 투과도는 증가하였다. Opposing reactants film deposition 방법으로 제조된 실리카 막은 비교적 안정성은 높으나 수소의 투과도가 낮은 반면, one-sided film deposition 방법을 사용하면 수소의 투과도는 높으나 안정성이 낮은 막이 얻어졌다. 이러한 실리카 막은 고온에서의 기체분리 및 분리막 반응기에 응용하기 위하여는 높은 선택적투과성 및 안정성이 요구되며 막 제조 조건 및 방법이 최적화되어야 함을 알 수 있었다.

• 한국수소및신에너지학회논문집
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• 제21권5호
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• pp.425-434
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• 2010

Integrated gasification combined cycle (IGCC) is an efficient and environment-friendly power generation system which is capable of burning low-ranked coals and other renewable resources such as biofuels, petcokes and residues. In this study some process modeling on a conceptual entrained flow gasifier was conducted using the ASPEN Plus process simulator. This model is composed of three major steps; initial coal pyrolysis, combustion of volatile components, and gasification of char particles. One of the purposes of this study is to develop an effective and versatile simulation model applicable to numerous configurations of coal gasification systems. Our model does not depend on the hypothesis of chemical equilibrium as it can trace the exact reaction kinetics and incorporate the residence time calculation of solid particles in the reactors. Comparisons with previously reported models and experimental results also showed that the predictions by our model were pretty reasonable in estimating the products and the conditions of gasification processes. Verification of the accuracy of our model was mainly based upon how closely it predicts the syngas composition in the gasifier outlet. Lastly the effects of change oxygen are studied by sensitivity analysis using the developed model.