• Advances in Automotive Engineering
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• 제1권1호
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• pp.61-77
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• 2018

In this study, the effects of sitting position of the driver on the whiplash neck injury have been analyzed experimentally by using hybrid III series 50 percentile male crash test dummy. A testing platform consisting of vehicle ground, driver foot rest, driver seat and a 3-point seatbelt has been prepared. This testing platform and the instrumented crash test dummy are prepared for tests according to the Euro NCAP whiplash testing protocol. The prepared test set-up has been exposed to 3 different acceleration-time loading curves defined in the Euro NCAP whiplash testing protocol by performing sled tests. 9 different sled tests have been performed with the combinations of 3 different seating positions of the crash test dummy and 3 different acceleration-time loading curves. The sensor data obtained from the crash test dummy and high-speed videos taken are analyzed according to the injury assessments criteria defined in the Euro NCAP whiplash testing protocol and the criticality of the whiplash injury is defined. It is seen that the backset distance of the driver head with the headrest and the height difference of the top of the head of the driver with the headrest have a great importance on whiplash injuries.

• 대한안전경영과학회지
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• 제16권4호
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• pp.11-19
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• 2014

Modern systems development becomes more and more complicated due to the need on the ever-increasing capability of the systems. In addition to the complexity issue, safety concern is also increasing since the malfunctions of the systems under development may result in the accidents in both the test and evaluation phase and the operation phase. Those accidents can cause disastrous damages if explosiveness gets involved therein such as in weapon systems development. The subject of this paper is on how to incorporate safety requirements in the design of safety-critical systems. As an approach, a useful system structure using the method of design structure matrix (DSM) is studied while reflecting the need on systems safety. Specifically, the effects of system components failure are analyzed and numerically modeled first. Also, the system components are identified and their interfaces are represented using a component DSM. Combining the results of the failure analysis and the component DSM leads to a modified DSM. By rearranging the resultant DSM, a modular structure is derived with safety requirements incorporated. As a case study, application of the approach is also discussed in the development of a military UAV plane.

• 한국가스학회지
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• 제22권2호
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• pp.1-20
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• 2018

Despite many measures, still from time to time catastrophic events occur, even after reviewing potential scenarios with HAZID tools. Therefore, it is evident that in order to prevent such events, answering the question: "What can go wrong?" requires more enhanced HAZID tools. Recently, new system based approaches have been proposed, such as STPA (system-theoretic process analysis) and Blended Hazid, but for the time being for several reasons their availability for general use is very limited. However, by making use of available advanced software and technology, traditional HAZID tools can still be improved in degree of completeness of identifying possible hazards and in work time efficiency. The new HAZID methodology proposed here, the Data-based semi-Automatic HAZard IDentification (DAHAZID), seeks to identify possible scenarios with a semi-automated system approach. Based on the two traditional HAZID tools, Hazard Operability (HAZOP) Study and Failure Modes, Effects, and Criticality Analysis (FMECA), the new method will minimize the limitations of each method. This will occur by means of a thorough systematic preparation before the tools are applied. Rather than depending on reading drawings to obtain connectivity information of process system equipment elements, this research is generating and presenting in prepopulated work sheets linked components together with all required information and space to note HAZID results. Next, this method can be integrated with proper guidelines regarding process safer design and hazard analysis. To examine its usefulness, the method will be applied to a case study.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2725-2732
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• 2020

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of ¹⁶O, ⁹Be, ²³⁵U, ²³⁸U, and S(𝛼,𝛽) of ¹H in H₂O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1881-1895
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• 2020

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

• Nuclear Engineering and Technology
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• 제21권4호
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• pp.245-258
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• 1989

열중성자로의 핵계산을 위한 69군 단면적 라이브러리를 생산하였다. 기본 평가핵자료로는 IAEA Nuclear Data Section에서 수집된 자료가, 그리고 이를 처리하여 군정수화 하는데는 NJOY코드가 이용되었다. 새로이 마련된 라이브러리의 유용성을 검증하기 위해 각기 산화우라늄과 금속 우라늄 연료로 구성된 임계실험치를 WIMS-KAERI 코드로 계산된 결과와 비교, 검토하였다. 총 88임계결과에 대해 평균 $K_{eff}$ 값 0.9997, 그리고 표준 편차 0.69%를 보였다. PWR 연료의 연소결과로 얻어진 우라늄과 플루토늄 생성량에 대한 평가에서도 전반적으로 좋은 결과를 얻었다.다.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

• 방사성폐기물학회지
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• 제4권1호
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• pp.41-50
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• 2006

본 연구에서는 사전연구로부터 사용후핵연료의 처분용기 원형모델로 제안된 처분용기의 전체 크기와 배열을 평가하기 위하여 일련의 공학적 분석을 수행하였다. 그러한 노력의 결과 용기 내부 저장통의 배열형태와 외곽쉘과 상하부 뚜껑의 두께와 같은 새로운 설계변수를 도출하였다. 공학적 분석 작업에는 처분용기의 기계구조 해석 결과를 근거로 도출된 용기의 규격자료에 대한 방사선 안전성 측면에서의 타당성을 검토하기 위하여 방사선차폐 해석과 핵 임계 해석 등이 수행되었다. 처분용기 내부 삽입체의 직경 변화에 따른 구조안정성 해석 결과에 따르면, 직경 102cm 일 때 극한 외압조건은 물론 정상적인 외압조건 하에서도 최대 Von Mises 응력이 안전계수 2.0을 만족하는 것으로 나타났다. 이 경우에서도 핵 임계 및 방사선차폐 해석 결과 안전기준치를 만족시키며, 무게는 20톤 가량 줄어드는 효과가 있는 것으로 나타났다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 추계 학술대회 논문집
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• pp.153-162
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• 2005

본 연구에서는 사전연구로부터 사용후 핵연료의 처분용기 원형모델로 제안된 처분용기의 전체 크기와 배열을 평가하기 위하여 일련의 공학적 분석을 수행하였다. 그러한 노력의 결과 용기 내부 저장통의 배열형태와 외곽쉘과 상하부뚜껑의 두께와 같은 새로운 설계변수를 도출하였다. 공학적 분석 작업에는 처분용기의 기계구조 해석 결과를 근거로 도출된 용기의 규격자료에 대한 방사선 안전성 측면에서의 타당성을 검토하기 위하여 방사선차폐 해석과 핵임계 해석 등이 수행되었다. 처분용기 내부 삽입체의 직경변화에 따른 구조안정성 해석 결과에 따르면, 직경 102cm일 때 극한 외압조건은 물론 정상적인 외압조건 하에서도 최대 Von Mises 응력이 안전계수 2.0을 만족하는 것으로 나타났다. 이 경우에도 핵임계 및 방사선차폐 해석 결과 안전기준치를 만족시키며, 무게는 20톤 가량 줄어드는 효과가 있는 것으로 나타났다.

• 지질공학
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• 제17권3호
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• pp.339-350
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• 2007

암반에 발달하고 있는 단층이나 절리와 같은 불연속면뿐만 아니라 점토의 함량비, 팽창성 점토물질의 협재, 배수특성(drainage) 등은 암반의 붕괴여부를 결정하는 중요한 인자들이다. 특히 점토광물의 성분은 강우에 의한 암반붕괴를 예측할 수 있는 중요한 지표가 될 수 있다. 최근 사면이나 터널의 설계에서도 그 중요성이 점차 증가하는 추세이다. 본 연구는 니질천매암과 사질천매암이 호층을 이루고 있는 OO 터널의 선진시추코어(horizontal boring core)를 이용하여 단층발달에 따른 광물성분의 변화 및 이들이 불연속면의 역학성에 미치는 영향을 파악하고, 암석 구성물질과 파생된 점토광물을 비교하여 암반의 불안정성을 추정하기 위해 수행되었다. 연구방법은 시추코어 중에서 단층의 영향을 받은 구간과 비교적 신선한 구간에서 시료를 채취하여 박편을 제작하여 관찰하였고, 점토광물의 성분 및 함량을 분석하였다. 그리고 야외조사와 실내 시험으로 단층물질의 강도정수를 구하였다.