• Nuclear Engineering and Technology
• /
• 제42권5호
• /
• pp.474-499
• /
• 2010

This paper presents the mostortant developments implemented in the APOLLO2 spectral code since its last general presentation at the 1999 M&C conference in Madrid. APOLLO2 has been provided with new capabilities in the domain of cross section self-shielding, including mixture effects and transfer matrix self-shielding, new or improved flux solvers (CPM for RZ geometry, heterogeneous cells for short MOC and the linear-surface scheme for long MOC), improved acceleration techniques ($DP_1$), that are also applied to thermal and external iterations, and a number of sophisticated modules and tools to help user calculations. The method of characteristics, which took over the collision probability method as the main flux solver of the code, allows for whole core two-dimensional heterogeneous calculations. A flux reconstruction technique leads to fast albeit accurate solutions used for industrial applications. The APOLLO2 code has been integrated (APOLLO2-A) within the $ARCADIA^{(R)}$ reactor code system of AREVA as cross section generator for PWR and BWR fuel assemblies. APOLLO2 is also extensively used by Electricite de France within its reactor calculation chain. A number of numerical examples are presented to illustrate APOLLO2 accuracy by comparison to Monte Carlo reference calculations. Results of the validation program are compared to the measured values on power plants and critical experiments.

• 터널과지하공간
• /
• 제21권1호
• /
• pp.20-32
• /
• 2011

개착과정에서 붕락현상이 발생한 암반사면의 거동양상을 분석하였다. 개착사면 암반의 특성 분석을 위하여 시추작업을 수행하였으며, BIPS 영상 분석을 통하여 불연속면의 방향성과 시추공 내의 출현 위치를 측정하였다. 회수된 코어를 관찰하여 암반 내부의 구조적 취약성을 고찰하였다. 절리면에 협재된 팽윤성 점토광물의 조성을 X-선 회절분석을 통하여 조사하였으며, 직접전단실험을 수행하여 절리면 전단강도를 측정하였다. 사면파괴를 유발시킬 수 있는 절리들의 위치 및 방향성을 고려한 블록해석을 수행하여 규모별 블록안정성을 산출하였다. 횡단면 해석기법을 이용하여 개착과정 중에 발생할 수 있는 지반거동 양상을 분석하였으며, 안정성 확보를 위하여 소요되는 보강량을 산정하였다. 보강작업이 수행된 개착사면의 추가적인 거동양상을 계측 해석을 통하여 고찰하였다.

• 대한건축학회논문집:구조계
• /
• 제35권7호
• /
• pp.171-178
• /
• 2019

Auto Climbing Formwork System (ACS) for construction of high-rise building is a construction method for automatically lifting the formwork system supported by the anchor on the pre-constructed concrete wall. It has excellent construction speed and quality, but it has the possibility of structural failure depending on the quality of concrete and also has low economical efficiency due to the use of foreign technology. In order to overcome these problems, this study conducted an optimum design for the development of a new concept of Corner Supported Auto Climbing System (CS-ACS) in conjunction with the development of corner steel-reinforced concrete core wall system. For the design the formwork system, the basic module and structural member compositions were planned, and the structural analysis program was used to analyze the optimum member's cross section and spacing. As a result, the horizontal displacement and the stress of the horizontal members were influenced by the spacing more than the cross-section of the member. On the other hand, vertical members did not affect the displacement and stress of the formwork system. The form tie was very effective in controlling the displacement when adjusting the spacing of the horizontal members, but when the spacing of the form tie is more than 1,500mm, it is analyzed that form tie is yielding in basic module. When the span of the formwork system is more than 30m, it is analyzed that the basic module needs to be changed because of the increase of overall displacement.

• Nuclear Engineering and Technology
• /
• 제41권5호
• /
• pp.729-738
• /
• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• 한국추진공학회:학술대회논문집
• /
• 한국추진공학회 2008년 영문 학술대회
• /
• pp.115-121
• /
• 2008

A core engine for pre-cooled turbojet engines is designed and its component performances are examined both by CFD analyses and experiments. The engine is designed for a flight demonstration of precooled turbojet engine cycle. The engine uses gas hydrogen as fuel. The external boundary including measurement devices is set within $23cm{\times}23cm$ of rectangular cross section, in order to install the engine downstream of the air intake. The rotation speed is 80000 rpm at design point. Mixed flow compressor is selected to attain high pressure ratio and small diameter by single stage. Reverse type main combustor is selected to reduce the engine diameter and the rotating shaft length. The temperature at main combustor is determined by the temperature limit of non-cooled turbine. High loading turbine is designed to attain high pressure ratio by single stage. The firing test of the core engine is conducted using components of small pre-cooled turbojet engine. Gas hydrogen is injected into the main burner and hot gas is generated to drive the turbine. Air flow rate of the compressor can be modulated by a variable geometry exhaust nozzle, which is connected downstream of the core engine. As a result, 75% rotation speed is attained without hazardous vibration and heat damage. Aerodynamic performances of both compressor and turbine are obtained and evaluated independently.

• Nuclear Engineering and Technology
• /
• 제46권3호
• /
• pp.313-342
• /
• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• 한국구조물진단유지관리공학회 논문집
• /
• 제25권6호
• /
• pp.131-142
• /
• 2021

노후 교각은 내진설계가 적용되지 않아 소성힌지구역에 겹침이음이 대다수 존재한다. 철근부식은 철근 단면적 감소 및 겹침이음부의 거동저하를 유발하여 교각의 내진성능을 저하시킨다. 본 연구에서는 이러한 노후교각의 특성에 따라 철근부식, 겹침이음, 내진설계 및 내진 보강 여부를 고려하여 실험체를 설계 및 제작하고 실험을 통해 그 영향을 조사하였다. 실험결과, 겹침이음 또는 철근부식은 변위연성도를 감소시킨다. 내진설계 상세 또는 강판 내진보강을 적용하면 충분한 변위연성도가 확보됨을 확인하였다. 모든 비내진실험체는 소성힌지구역 내의 횡철근 겹침이음부의 풀림으로 인해 주철근 좌굴과 심부콘크리트 압축파쇄가 발생하였다. 내진설계된 실험체는 철근부식에 의한 소성힌지구역 내 횡철근의 단면감소와 갈고리 풀림에 의해 주철근 좌굴 및 심부콘크리트 압축파쇄가 발생하였다.

• 한국의류학회지
• /
• 제19권6호
• /
• pp.946-954
• /
• 1995

The purpose of this study is to modify the hydrophobic property and dyeability of polyethylene terephthalate fiber. Methacrylic acid (2nA) was graftpolymerized with benzoyl peroxide (BPO) as initiator onto polyethylene terephthalate fabrics. The results were as follow; 1. Graft-polymerization exhibited maximum graft ratio at a temperature of 100"C. 2. The polymer was gradually grafted in great amount to the surface of MA-g-PET as graft ration increase; with the cross-section examination of MA-g-PET, it was discovered that graft-polymeriation had also taken place inside the textile core. 3. Dyes absorption of basic dyes and disperse dyes was improved as craft ratio increase; with resistance to laundering, the former showed grade 3-4 and the latter showed grade 5.de 5.

• 태양에너지
• /
• 제9권3호
• /
• pp.64-72
• /
• 1989

In this study, the computer simulation for the heat transfer in pulse combustion water heater is performed. The attention is focused to the effects of the installation of corebuster in the flue tube on heat transfer. The energy equations are established for both wall and gas side in the combustion chamber, flue way, exhaust chamber and muffler, and the numerical calculation is executed. Zone method takes longer computer calculation time compared with semi-zone method. Semi-zone method is chosen for numerical calculation. As a result of this study, it is found that the installation of the core buster in flue tube increases total heat transfer. It is also found that the total heat transfer is increased with the increasing of the ratio of the cross section area of corebuster to that of the flue tube. However, the heat transfer effect is negligible for the area ratio above 0.5.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
• /
• pp.771-776
• /
• 1998

TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Vnit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) far all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library. BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The makimum fast neutron nun calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effgctive full power days is 1.784x10$^{18}$ n/$\textrm{cm}^2$. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60cm below the midplane at zero degree.