• Nuclear Engineering and Technology
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• 제49권4호
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• pp.792-800
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• 2017

The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma $({\gamma})-rays$. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, ${\gamma}-ray$ shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV-15 MeV) and optical thickness (OT). The study was performed by computing various ${\gamma}-ray$ shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

• 센서학회지
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• 제25권3호
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• pp.229-234
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• 2016

Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the vanadium (V) SPND has been being developed to be used in OPR1000 nuclear power plants. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina insulator with a cylindrical geometry. An MCNP code was used to simulate some factors (neutron self-shielding factor and beta escape probability from the emitter) and space charge effect of an insulator necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND and contribute to the development of TMI (Top-mount In-core Instrumentation) which might be used in the SMART and SMR.

• 한국융합학회논문지
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• 제11권11호
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• pp.89-94
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• 2020

자기공명영상 고주파를 차단할 수 있는 금속물질 중 자장에 영향을 받지 않으며 가격이 저렴하고 주위에서 쉽게 구할 수 있는 알루미늄 호일을 이용하여 차폐체를 제작하였다. Phantom 실험 결과 over-sampling 기법을 적용한 경우 적용 전보다 aliasing artifact가 약 94% 감소하였고, 알루미늄 차폐체를 적용한 경우도 적용 전 보다 약 92% 감소하였다. 그리고 scan time도 over-sampling 기법의 경우 적용 전보다 약 3배이상 증가한 반면, 알루미늄 차폐체의 경우는 적용 전과 변화가 없는 것으로 나타났다. 따라서 알루미늄 호일을 이용한 차폐체도 기존의 흡수재 및 차폐체들과 마찬가지로 scan time의 증가 없이 aliasing artifact를 효율적으로 제거할 수 있는 것을 확인하였다.

• 콘크리트학회논문집
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• 제29권2호
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• pp.169-177
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• 2017

북한은 최근 핵무기의 일반적인 열, 폭풍, 방사능 피해가 아닌 전자 장비를 무력화시키기 위한 고고도 전자기파 탄을 개발 중인 것으로 예측되고 있다. 현재 군용 목적으로 사용되고 있는 HEMP 차폐 시설 중 차폐 판의 경우 전자파 차폐 효과가 뛰어난 금속 판이 사용되고 있으나 이러한 금속판들은 차폐 시설 제작 시 용접 부위에서의 전자파 유입 가능성 등 시공상의 어려움과 높은 비용이 문제시 되고 있는 실정이다. 이에 본 연구에서 차폐 시설을 따로 구축하지 않고 콘크리트 구조물 자체로써 전자파 차폐 효과를 확보하기 위하여 콘크리트 실험체에 전기전도성이 높은 재료를 혼입하였다. 또한, 실험체 중 가장 높은 차폐효과를 보인 2가지 수준과, 가장 낮은 차폐 효과를 보인 2가지 수준에 $100{\mu}m$ 아연-알루미늄 합금 금속용사 피막을 적용하였다. 실험 결과 전기전도성이 높은 재료를 혼입한 실험체는 MIL-STD-118-125-1 규격 최소 차폐 기준을 만족하지 못하였으나, 금속용사 피막을 적용한 실험체에서는 모두 최소 차폐 기준을 만족하였다. 결론적으로, $100{\mu}m$ 아연-알루미늄 합금 금속용사 피막이 HEMP 차폐에 높은 효율성을 가지고 있다고 판단된다.

• Journal of Welding and Joining
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• 제18권6호
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• pp.68-73
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• 2000

In case of overalyered wrapping roll, beadmark shape appear at wrapping roll surface due to irregular wear between weld bead. Irregular wear of this is caused by difference of hardness between weld bead. This study aims at investigating which matrix is good for removing of beadmark at wrapping roll surface. So, we make specimen with martensitic matrix and austenitic matrix. The overlayered alloys were deposited 4 times on a SS41 steel plate using self-shielding flux cored arc welding method. Difference of hardness between weld bead of specimen with matrix of martensite was higher than specimen with matrix of austenite both as-welded and after heat treatment. Therefore, austenitic matrix is between than martensitic matrix for removing of beadmark of wrapping roll surface.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.49.2-49
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• 2002

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

• 한국방사선학회논문지
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• 제13권1호
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• pp.45-53
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• 2019

본 연구는 의료용 선형가속기 시설을 차폐하는 콘크리트에 대한 중성자 방사화 연구로써, 일반 콘크리트와 저 방사화 콘크리트를 비교 분석하였다. 실험 방법은 MCNPX (Ver. 2.5.0)와 FISPACT-2010를 사용하여 모의실험을 진행하여, 광자선과 중성자선에 대한 차폐능을 산정하고 중성자 방사화 평가를 진행하였다. 그 결과 차폐능은 일반 콘크리트에서 20~50 cm 효율적이였으며, 방사화 평가의 경우 저 방사화 콘크리트에서 방사능이 낮게 계산되었으나, 모두 자체처분허용 농도를 초과하지 않는 수준으로 산정되었다. 이를 종합적으로 분석한 결과 일반 콘크리트를 사용하는 것이 효율적인 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1269-1286
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• 2017

This paper presents a newly developed resonance self-shielding method based on Tone's method in $APOLLO3^{(R)}$ for fast and thermal reactor calculations. The new method is based on simplified models, the narrow resonance approximation for the slowing down source and Tone's approximation for group collision probability matrix. It utilizes mathematical probability tables as quadrature formulas in calculating effective cross-sections. Numerical results for the ZPPR drawer calculations in 1,968 groups show that, in the case of the double-column fuel drawer, Tone's method gives equivalent precision to the subgroup method while markedly reducing the total number of collision probability matrix calculations and hence the central processing unit time. In the case of a single-column fuel drawer with the presence of a uranium metal material, Tone's method obtains less precise results than those of the subgroup method due to less precise heterogeneous-homogeneous equivalence. The same options are also applied to PWR UOX, MOX, and Gd cells using the SHEM 361-group library, with the objective of analyzing whether this energy mesh might be suitable for the application of this methodology to thermal systems. The numerical results show that comparable precision is reached with both Tone's and the subgroup methods, with the satisfactory representation of intrapellet spatial effects.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1287-1300
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• 2017

Based on the observation that ignoring the angle dependency of multigroup resonance cross sections within a fuel pellet would result in nontrivial underestimation of the spatial self-shielding of flux, a parametrized spectral superhomogenization (SPH) factor library (PSSL) method is developed as a practical means of resolving the problem. Region-wise spectral SPH factors are calculated by the normal and transport corrected SPH iterations after ultrafine group slowing down calculations over various light water reactor pin-cell configurations. The parametrization is done with fuel temperature, U-238 number density, fuel radius, moderator source represented by ${\Sigma}_{mod}V_{mod}$, and the number density ratio of resonance nuclides to that of U-238 in a form of resonance interference correction factors. The parametrization is successful in that the root mean square errors of the interpolated SPH factors over the fuel regions of various pin-cells are within 0.1%. The improvement in reactivity error of the PSSL method is shown to be superior to that by the original SPH method in that the reactivity bias of -200 pcm to -300 pcm vanishes almost completely. It is demonstrated that the environment effect takes only about 4% in the reactivity improvement so that the pin-cell based PSSL method is effective in the assembly problems.