• 한국재료학회:학술대회논문집
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• 한국재료학회 2011년도 춘계학술발표대회
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• pp.15-15
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• 2011

As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.

• 한국방사선학회논문지
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• 제9권7호
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• pp.459-465
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• 2015

PET/CT에서 사용되는 511 keV ${\gamma}$선의 납 차폐체 사용 유 무에 따른 에너지 흡수 분포를 몬테카를로 모의 모사를 통해 평가하였다. 실험은 ICRU Slab 팬텀을 이용하여 깊이에 따라 피부표면(0.07), 수정체(3), 심부(10)에 대해 실험을 진행하였으며, 납 두께에 따른 에너지 흡수 분포 차이와 납과 팬텀의 거리에 따른 공기층의 영향에 대해 분석 하였다. 그 결과 납 차폐체 사용 시 산란전자선에 의해 피부표면에 에너지 흡수 분포가 높게 나타났다. 산란전자선선은 납과 팬텀 사이의 거리가 증가함에 따라 점차 제거되었으며, 0.25 mm 납 차폐체 사용 시 9 cm 이상의 공기층이 있어야 피부표면의 도달하는 산란전자선의 영향을 방지 할 수 있었다. 또한 0.5 mm의 납 차폐체 사용 시 1 cm 이상의 공기층이 있어야 피부표면에 도달하는 산란전자선의 영향을 방지 할 수 있었으며, 공기층을 고려하지 않을 경우 0.75 mm이상의 납 두께를 사용하여야 피부표면의 산란전자선의 영향을 방지 할 수 있다.

• Nuclear Engineering and Technology
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• 제40권1호
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• pp.93-98
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• 2008

Several adhesive tapes have been studied in terms of their suitability for securing gold wires into positions for neutron flux measurements in the reactor core and irradiation facilities surrounding the core of the Open Pool Australian Light water (OPAL) reactor. Gamma ray spectrometry has been performed on each irradiated tape in order to identify and quantify activated components. Numerous metallic impurities have been identified in all tapes. Calculations relating to both the effective neutron shielding properties of the tapes and the error in measurement of the $^{198}Au$ activity caused by superfluous activity due to residual tape have been made. The most important identified effects were the prolonged cooling times required before safe enough levels of radioactivity to allow handling were reached, and extra activity caused by residual tape when measured with an ionisation chamber. Knowledge of the most suitable tape can allow a minimal contribution due to these effects, and the use of gamma spectrometry in preference to ionisation chamber measurements of the flux wires is shown to make all systematic errors due to the tape completely negligible.

• Nuclear Engineering and Technology
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• 제54권2호
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• pp.701-708
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• 2022

The mass attenuation coefficients (μ_m) of polyethylene glycol (PEG) of different molecular weights (1000-200,000) were measured using single-beam photon transmission. The X-ray fluorescent (XRF) photons from Zinc (Zn), Zirconium (Zr), Molybdenum (Mo), Silver (Ag) and Cadmium (Cd) targets were used to determine the attenuation of gamma radiation of energy range between 8.67 and 23.19 keV in PEG samples. The results were compared to theoretical values using XCOM and Monte Carlo simulation using Geant4 toolkit which was developed to validate the experiment at those certain energies. The mass attenuation coefficients were then used to compute the effective atomic numbers, electron density and half value layers for the studied samples. The outcomes showed good agreement between experimental and simulated results with those calculated theoretically by XCOM within 5% deviation. The PEG 1000 sample showed slightly higher μ_m value compared with the other samples. The dependence of the photon energy and PEG composition on the values of μ_m and HVL were investigated and discussed. In addition, the values of Z_eff and N_eff for all PEG samples behaved similarly in the given photon energy range, and they decreased as the photon energy increased.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.249-253
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• 2001

Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of $^{24}Na$ in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of $^{24}Na$ is approximately 50Bq for 10miniute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7 $[(Bq^{24}Na/g^{23}Na)/mGy]$.

• 한국방사선학회논문지
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• 제11권4호
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• pp.189-195
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• 2017

방사선 특히, 엑스선 또는 감마선으로부터 인체를 보호하기 위해 납(Pb)으로 된 보호 장구를 광범위하게 사용해왔다. 최근 납 중독 및 환경오염의 문제로 납을 대신하는 무연 방사선 차폐재의 개발이 활발히 이루어지고 있다. 차폐재의 성능 확보를 위해서는 제작 및 평가의 순환 사이클을 반복하게 된다. 본 연구는 실제 무연 방사선 차폐소재의 제작에 앞서 차폐재의 성능을 몬테카를로 전산모사를 통해 확인함으로써 가능한 차폐소재의 조합을 연구하였다. 방사선 차폐소재의 평가에 사용되는 조건으로 엑스선관을 Geant4를 이용하여 전산모사하고 획득된 광자 스펙트럼을 이용하여 텅스텐과 비스무스의 조합에 따른 차폐소재의 성능을 평가하였다. 차폐소재의 공극에 따른 성능 저하도 평가하였다. 방사선 차폐 소재 개발 시 공극률을 줄이는 것이 중요한 인자라는 것을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3913-3923
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• 2023

Todays, medium energy resolution detectors are preferably used in radioisotope identification devices(RID) in nuclear and radioactive material categorization. However, there is still a need to develop or enhance « automated identifiers » for the useful RID algorithms. To decide whether any material is SNM or NORM, a key parameter is the better energy resolution of the detector. Although masking, shielding and gain shift/stabilization and other affecting parameters on site are also important for successful operations, the suitability of the RID algorithm is also a critical point to enhance the identification reliability while extracting the features from the spectral analysis. In this study, a RID algorithm based on Bayesian statistical method has been modified for medium energy resolution detectors and applied to the uranium gamma-ray spectra taken by a LaBr3:Ce detector. The present Bayesian RID algorithm covers up to 2000 keV energy range. It uses the peak centroids, the peak areas from the measured gamma-ray spectra. The extraction features are derived from the peak-based Bayesian classifiers to estimate a posterior probability for each isotope in the ANSI library. The program operations were tested under a MATLAB platform. The present peak based Bayesian RID algorithm was validated by using single isotopes(²⁴¹Am, ⁵⁷Co, ¹³⁷Cs, ⁵⁴Mn, ⁶⁰Co), and then applied to five standard nuclear materials(0.32-4.51% at.²³⁵U), as well as natural U- and Th-ores. The ID performance of the RID algorithm was quantified in terms of F-score for each isotope. The posterior probability is calculated to be 54.5-74.4% for ²³⁸U and 4.7-10.5% for ²³⁵U in EC-NRM171 uranium materials. For the case of the more complex gamma-ray spectra from CRMs, the total scoring (ST) method was preferred for its ID performance evaluation. It was shown that the present peak based Bayesian RID algorithm can be applied to identify ²³⁵U and ²³⁸U isotopes in LEU or natural U-Th samples if a medium energy resolution detector is was in the measurements.

• 한국의학물리학회지:의학물리
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• 제22권1호
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• pp.3-11
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• 2011

본 연구에서는 제작한 수직형 다엽 콜리메이터를 이용하여 방사선치료에 사용되는 Co-60 감마선 및 6 MV 엑스선의 조사면 크기와 모양을 결정하고 동일한 모양 및 크기의 조사면을 납차폐체로 결정하여 방사선 조사면 내 선량분포 특성을 상호 분석하여 수직형 다엽 콜리메이터의 방사선 조사면 크기 결정에 관한 유용성을 평가하였다. 이를 위해 이온전리함, 유리선량계, 방사선크로믹 필름을 사용하여 선량측정 실험을 수행하였다. Co-60 감마선과 6 MV 엑스선에 대하여 기준조사면의 이온전리함 측정결과 수직형 다엽 콜리메이터의 빔 중심축 선량값이 납차폐체의 선량값보다 각각 5.1%, 4.2% 높게 측정되었다. 그리고 Co-60 감마선에 대한 4개 조사면(기준 조사면, 원형, 삼각형, 십자형)의 유리선량계 측정 결과는 수직형 다엽 콜리메이터의 선량값이 납차폐체의 선량값보다 각각 2.2%, 7.8%, 7.2%, 4.0% 높게 측정되었고, 6 MV 엑스선에 대하여는 수직형 다엽 콜리메이터의 선량값이 납차폐체의 선량값보다 각각 6.7%, 6.2%, 3.8%, 6.2% 높게 측정되었다. 방사선크로믹 필름에서 차폐체의 선량분포곡선 중 최대선량의 80%에서 20%까지의 거리를 나타내는 반음영 크기는 모든 조사면에서 수직형 다엽 콜리메이터의 반음영 크기가 납차폐체보다 Co-60의 경우 2.0~3.5 mm, 6 MV 엑스선의 경우 0.5~1.0 mm 작게 나타났으며 이는 제작한 수직형 다엽 콜리메이터가 임상에 사용되었을 때 반음영의 크기를 납차폐체보다 줄일 수 있음으로써 치료 조사면적 결정시 차폐물의 반음영으로 생기는 방사선치료체적(Treatment Volume, TV)을 최소화시킬 수 있는 장점이 있으리라 판단된다. 아울러 2차원 및 3차원 방사선치료 시 본 다엽 콜리메이터를 이용하여 다양한 방사선치료 조사면을 간편하게 결정하여 사용할 수 있으리라 생각된다.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3808-3815
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• 2021

In this paper, the EPICS2017 photoatomic database was used to evaluate the photon mass attenuation coefficients and buildup factors of soils collected at different depths in the Philippine islands. The extraction and interpolation of the library was accomplished at the recommended linear-linear scales to obtain the incoherent and total cross section and mass attenuation coefficient. The buildup factors were evaluated using the G-P fitting method in ANSI/ANS-6.4.3. An agreement was achieved between XCOM, MCNP5, and EPICS2017 for the calculated mass attenuation coefficient values. The buildup factors were reported at several penetration depths within the standard energy grid. The highest values of both buildup factor classifications were found in the energy range between 100 and 400 keV where incoherent scattering interaction probabilities are predominant, and least at the region of predominant photoionization events. The buildup factors were examined as a function of different soil silica contents. The soil samples with larger silica concentrations were found to have higher buildup factor values and hence lower shielding characteristics, while conversely, those with the least silica contents have increased shielding characteristics brought by the increased proportions of the abundant heavier oxides.

• 방사성폐기물학회지
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• 제9권2호
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• pp.73-80
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• 2011

한국방사성폐기물관리공단 주관 하에 개념 설계된, 연소도이득효과 적용 대용량 수송용기에 대해 방사선 차폐 안전성을 평가하였으며 여러 방사선원들이 수송용기 주변 선량률 분포에 미치는 영향을 분석하였다. 가능한 모든 방사선원(중성자선원, 감마선원, 방사화선원)들을 고려하였으며 보수적인 가상의 핵연료(너비: WH 17 RFA, 축방향: CE Type)를 선정, 실제 상황과 동일한 조건이 되도록 계산모델을 구축하였다. 모든 조건(정상 및 가상사고 조건)에서 표면선량률과 외부선량률이 법적기준치를 만족하고 있었으며 축방향 높이에 따라 각 선원들의 기여도가 변하고 있었지만 정상조건에서의 최대 표면선량률과 외부선량률은 방사화선원에 의한 영향이 가장 높은 것으로 확인되었다. 가상사고 조건에서는, 중성자선원의 선량률 기여도가 대략 90%에 달하고 있었으나 수송용기 끝단에서는 방사화선원에 의한 선량률이 급격하게 상승함에 따라 BUC 적용 수송용기의 방사선 차폐해석시 충분히 보수적으로 해석되도록 방사화선원을 정밀하게 분석하여 설정하여야 할 것으로 판단되었다.