• Nuclear Engineering and Technology
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• 제3권4호
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• pp.185-197
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• 1971

1950년대의 미국 General Atomic사에서 열출력 100 kw로 설계, 제작하여 1962년 3월에 건조완료한 TRIGA Mark-II원자로는 1969년 7월에 250 kw로 출력 증강되었으나 방사선차폐는 보강되지 않았다. 본 논문에서의 계산에 의하면 출력 증강후 현재의 차폐물로도 중성자에 대하여는 확실히 안전하지만 Gamma선에 대해서는 위험하다는 것이 판명되었다. 원자로의 구조와 출입인 및 실험종사자들의 위치로 보아 차폐물의 안전도 검토는 수평방향에 한하였고, 또 정확을 기하기 위하여 중성자와 Gamma선의 투과문제를 나누어 검토하였다. 이를 근거로 하여 이론적인 측면에서 본 콘크리트의 보강을 요하는 두께도 산출하였다.

• Nuclear Engineering and Technology
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• 제41권3호
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• 한국산학기술학회논문지
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• 제6권1호
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• pp.54-57
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• 2005

반도체 소자 면적의 축소에 따라 중성자의 소프트 에러율은 집적회로 설계시 큰 문제점으로 대두되고 있다. 고전류 중성자 빔에 의한 가속 실험에서, 래치-업 현상은 소프트 에러 발생율의 정확한 예측을 방해하는 요소로 작용하고 있다. 본 연구는 SRAM 소자의 SER 가속 실험시 발생하는 래치-업에 대한 효과를 분석하였다. 2차원 소자 시뮬레이터를 이용한 시뮬레이션 환경하에서의 결과 깊은 p-well 구조의 기판이 이중 또는 삼중 well 구조에 비하여 양호한 래치-업 방지 효과를 나타내었다. 또한 접지에 대한 $V_{DD}$ 전력선까지의 거리를 최소화하는 것이 효과적인 설계 기법으로 평가되었다.

• 대한방사선기술학회지:방사선기술과학
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• 제43권5호
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• pp.389-395
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• 2020

The evaluation of radioactivated components of heavy-ion accelerator facilities affects the safety of radiation management and the exposure dose for workers. and this is an important issue when predicting the disposal cost of waste during maintenance and dismantling of accelerator facilities. In this study, the FLUKA code was used to simulate the proton treatment device nozzle and classify the radio-nuclides and total radioactivity generated by each component over a short period of time. The source term was evaluated using NIST reference beam data, and the neutron flux generated for each component was calculated using the evaluated beam data. Radioactive isotopes caused by generated neutrons were compared and evaluated using nuclide information from the International Radiation Protection Association and the Korea Radioisotope association. Most of the nuclides produced form of beta rays and electron capture, and short-lived nuclides dominated. However, In the case of ⁵⁴Mn, which is a radioactive product of iron, the effect of gamma rays should be considered. In the case of tritium generated from a material with a low atomic number, it is considered that handling care should be taken due to its long half-life.

• 분석과학
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• 제18권4호
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• pp.263-270
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• 2005

핵분열트랙은 우라늄, 플루토늄, 토륨 등의 방사성핵종 분석이나, 리튬이나 붕소와 같이 열중성자와 핵반응을 하는 원소분석에 매우 유용하다. 이 방법은 대기 중의 방사성핵종 분석을 통해 핵무기 실험을 검출한다거나, 우주공간의 중이온에 대한 선량측정이나 운석의 우주선 조사과정 추정, 원자력분야의 핵분열률 측정, 암석의 연대 및 역사를 파악하거나, tracer로 사용하는 등 여러 분야에 적용 가능하다. 현재 원자력연구소에서는 알파트랙기입법을 이용한 미량의 보론 분석이나, 핵분열트랙을 이용한 우라늄 등의 핵물질을 포함한 입자분석에 적용하고 있다. 본 총설에서는 트랙형성의 이론적 배경, 에칭된 트랙을 얻는 실제 과정 및 앞으로의 전망 등에 대해 살펴보았다.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.1-8
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• 2019

The integrity of the Reactor Pressure Vessel (RPV) is affected by the neutrons bombarding the vessel wall leading to embrittlement. This irradiation-induced embrittlement leads to reduction in the fracture toughness of RPV materials. This paper presents a comparative study of typical Optimized Power Reactor (OPR)1000 reactor pressure-temperature (P-T) limit curves using the pre-2006 American Society of Mechanical Engineers (ASME) editions used in the power plant and the current ASME edition of 2010. The current ASME Code utilizes critical reference stress intensity factor based on the lower bound of static, while the Pre-2006 ASME editions are based the critical reference stress intensity factor based on the lower bound of static, dynamic and crack arrest. Model-Based Systems Engineering approach was used to evaluate ASME Code Section XI Appendix G for generating the P-T limit curves. The results obtained from this analysis indicate decrease in conservatism in P-T limit curves constructed using the current 2017 ASME code, which can potentially increase operational flexibility and plant safety. Hence it is recommended to use ASME code edition after 2006 be used in all operating nuclear power plants (NPPs) to establish P-T limit curve.

• Journal of Radiation Protection and Research
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• 제36권4호
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• pp.174-182
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• 2011

With the reactor operation conditions - 4.3 wt% $^{235}U$ initial enrichment, burn-up 55,000 MWd/MTU, average power 34 MW/MTU for three periods burned time for 539.2 days per period and cooling time for 100 hours after shut down, to set up the condition to determine the minimum height (depth) of spent fuel storage pool to shut off the radiation out of the spent fuel storage pool and to store spent fuels safely, the dose rate on the specific position directed to the surface of spent fuel storage pool induced by the neutron and gamma-ray from spent fuels are evaluated. The length of spent fuel is 381 cm, and as the result of evaluation on each position from the top of spent fuel to the surface of spent fuel storage pool, it is difficult for neutrons from spent fuels to pass through the water layer of maximum 219 cm (600 cm from the floor of spent fuel storage pool) and 419 cm (800 cm from the floor of spent fuel storage pool) for gamma-ray. Therefore, neutron and gamma-ray from spent fuels can pass through below 419 cm (800 cm from the floor) water layer directed to the surface of spent fuel storage pool.

• The Korean Journal of Ceramics
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• 제6권3호
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• pp.240-244
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• 2000

The three-dimensional orientation distribution function of a conical shaped tungsten liner prepared by the thermo-mechanical forming process was analyzed by 1.525$\AA$ neutrons to carry out the Rietveld refinement. The pole figure data of three reflections, (110)(220) and (211) were measured. The orientation distribution functions for the normal and radial directions were calculated by the WIMV method. The inverse pole figures of the normal and radial directions were obtained from their orientation distribution functions. The Rietveld refinement was performed with the RIETAN program that was slightly modified for the description of preferred orientation effect. We could successfully do the Rietveld refinement of the strongly textured tungsten liner by applying the pole density of each reflection obtained from the inverse pole figure to the calculated diffraction pattern. The correction method of preferred orientation effect based on the inverse pole figures showed a good improvement over the semi-empirical texture correction based on the direct usage of simple empirical functions.

• Nuclear Engineering and Technology
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• 제38권4호
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• pp.327-342
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• 2006

Radiation therapy is an important part of cancer treatment in which cancer patients are treated using high-energy radiation such as x-rays, gamma rays, electrons, protons, and neutrons. Currently, about half of all cancer patients receive radiation treatment during their whole cancer care process. The goal of radiation therapy is to deliver the necessary radiation dose to cancer cells while minimizing dose to surrounding normal tissues. Success of radiation therapy highly relies on how accurately 1) identifies the target and 2) aim radiation beam to the target. Both tasks are strongly dependent of imaging technology and many imaging modalities have been applied for radiation therapy such as CT (Computed Tomography), MRI (Magnetic Resonant Image), and PET (Positron Emission Tomogaphy). Recently, many researchers have given significant amount of effort to develop and improve imaging techniques for radiation therapy to enhance the overall quality of patient care. For example, advances in medical imaging technology have initiated the development of the state of the art radiation therapy techniques such as intensity modulated radiation therapy (IMRT), gated radiation therapy, tomotherapy, and image guided radiation therapy (IGRT). Capability of determining the local tumor volume and location of the tumor has been significantly improved by applying single or multi-modality imaging fur static or dynamic target. The use of multi-modality imaging provides a more reliable tumor volume, eventually leading to a better definitive local control. Image registration technique is essential to fuse two different image modalities and has been In significant improvement. Imaging equipments and their common applications that are in active use and/or under development in radiation therapy are reviewed.

• Nuclear Engineering and Technology
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• 제10권3호
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• pp.145-151
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• 1978

D$_2$O에 대한 Butler의 scattering kernel에 관해 조사하였다. 이 kernel을 사용한 전자계산 코드 DEUKER를 개발하였으며 이를 이용하여 산란법칙, 미분산란 단면적 및 전산란 단면적을 계산하였다. D$_2$O 분자내의 어떤 두개 원자간의 산란 영향은 열에너지 영역에서의 산란 중성자 분포를 결정하는데 있어서 중요한 인자가 된다. 따라서 여러 입사 중성자 에너지에 따른 에너지 전달 산란 단면적에 관해 조사하여 이 kernel의 원자로 내의 중성자 분포 계산을 위한 사용 가능성을 연구하였다. 또한 본 연구를 통하여 D$_2$O와 $H_2O$에서의 중성자 산란 과장의 차이점을 밝혔다.