• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.268-273
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• 1997

사용후핵연료는 장기간 강한 방사선과 붕괴열이 방출된다. 따라서 사용후핵연료를 안전하게 운반하기 위하여 수송용기는 방사선차폐의 건전성, 격납경계의 유지 및 내부 붕괴열의 적절한 제거 등의 설계기준을 만족하도록 설계되어야 한다. 본 연구에서는 28개의 PWR 사용후핵연료집합체를 운반할 수 있는 KSC-28 수송용기의 적절한 열전달 특성을 갖는 copper 냉각핀 및 aluminum 전열판을 설정하였다. 또한, 정상수송조건 및 화재사고조건에 대한 열전달해석을 수행하여 수송용기의 열적 건전성을 평가하였고 여기에서 얻어진 온도를 열하중으로 고려하여 열응력해석을 수행함으로써 수송용기의 온도변화에 따른 구조적 건전성을 평가하였다.

• Journal of Radiation Protection and Research
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• v.19 no.1
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• pp.13-22
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• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

• Proceedings of the Korean Geotechical Society Conference
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• 2006.03a
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• pp.42-50
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• 2006

역들의 공간적 분포가 불균질하고 역의 크기가 큰 역암의 경우 암석 전체를 대표하는 물성치($E_m,\;c,\;\Phi$ 등) 구하기 위해서는 매우 큰 시험기기가 필요하다. 따라서 커다란 역을 포함하는 역암의 경우 직접 암석실내시험을 통한 물성치 산정은 현실적으로 거의 불가능하다. 이러한 문제를 극복하기 위하여 이 연구에서는 개별요소법을 이용하여 역암의 물성치를 산출하는 방법을 제안한다. 그 방법은다음과 같다. (1) 역암내의 역의 물성과 기질부의 물성을 각각 실내실험을 통하여 파악한 후 이들 (2) 두 물질의 거동양상을 구현할 수 있는 개별요소집합체의 개별요소간의 물성을 결정한다. (3) 역의 함량, 크기 모양 공간적 분포양상등의 역암 조직과 유사한 개별요소 수치해석시료를 만든 후, (4) 이를 수치 해석실험 (이축압축실험)에 사용한다. 이러한 수치해석실험을 통해 현재까지 만들어진 결과는 다음과 같다. 첫째, 역의 강도가 기질의 강도보다 높은 역암의 경우, 역의 양이 증가할수록 일축압축강도, 내부 마찰각, 점착력이 증가하지만 증가 양상은 선형이 아니다. 탄성계수의 경우 역의 양과 상관 없이 변화하지 않는다. 둘째, 역과 기질 사이 표면의 점착력이 약할 경우 이러한 표면에서 최초 미세 균열이 형성되기 시작하므로 이 점착력은 물성치를 산출하는 중요한 인자이다. 따라서, 향후 이에 대한 자세한 연구가 필요하다고 판단된다. 결론적으로,설계 또는 시공시 직접시험에 의한 물성치의 파악이 어려운 역암 또는 직접시험을 위해 대량의 시료를 필요로 하는 함력 미고결지층, 핵석층, 풍화암과 같은 시료의 물성치는 별도로 측정된 물성들 (예, 역과 기질)을 이용한 개별요소법을 통해 구할 수 있다.로 나타났다.TEX>, DIN/DIP비 표층수 $23.91\pm3.42$, 저층수 $23.43\pm3.38$이었으며, 전반적으로 해역별 수질기준 I등급 내지는 II등급을 유지하고 있었고, 공간적으로는 외해측으로 갈수록 외해수와 혼합 확산되어 양호한 수질을 나타내었다. 장기적인 변동특성은 세그룹으로 구분되어진다.기 실험결과 용출용매로 증류수와 해수를 이용했을 때, 제강 슬래그에서 용출되는 납, 구리, 카드뮴, 수은의 용출 경향의 차이를 확인할 수 있었고 이에 따라서, 납, 구리, 카드뮴의 용출 유해성은 낮기 때문에 해양구조물로의 제강슬래그 유효이용은 적합할 것으로 판단되었다.im80%$로 계산되었다. 열형광선량계로 측정된 방사선량은 각각 1.8, 1.2, 0.8, 1.2, 0.8 (70 cm 거리) cGy로 측정되었으며, 환자의 복부 표면에서의 서베이메터를 이용한 측정량은 10.9 mR/h였다. 차폐구조물의 사용 시 전체 치료 동안에 태아선량은 약 1 cGy 정도로 평가되었다. 결론 : AAPM Report No.50의 자료에 따르면, 임산부의 방사선 치료 시 태아의 방사선 피폭선량은 5 cGy 이하일 경우에 방사선 피폭에 따른 태아의 위험이 거의 없는 것으로 제시되고 있다. 본원에서 차폐 구조물을 설치하였을 경우에 측정된 태아선량은 약 1 cGy로 측정되었고, 고안된 차폐구조물은 태아에 도달하는 방사선량을 감소시키기에 적합한 설계임이 입증되었다. 아니라 일반종합병원에서도 CTX-M형 ESBL 생성 E. coli와 K. pneumoniae가 존재하며 확산 중임을 시사한다. 앞으로 CTX-M형 ESBL의 만연과 변종 CTX-M형 ESBL의 출연을 감시하기 위한

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.73-80
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• 2011

The radiation shielding analysis for a Burnup-credit (BUC) cask designed under the management of Korea Radioactive Waste Management Corporation (KRMC) was performed to examine the contribution of each radiation source affecting dose rate distribution around the cask. Various radiation sources, which contain neutron and gamma-ray sources placed in active fuel region and the activation source, and imaginary nuclear fuel were all considered in the MCNP calculation model to realistically simulate the actual situations. It was found that the maximum external and surface dose rates of the spent fuel cask were satisfied with the domestic standards both in normal and accident conditions. In normal condition, the radiation dose rate distribution around the cask was mainly influenced by activation source ($^{60}Co$ radioisotope); in another case, the neutron emitted in active fuel region contributed about 90% to external dose rate at 1m distance from side surface of the cask. Besides, the contribution level of activation source was dramatically increased to the dose rates in top and bottom regions of the cask. From this study, it was recognized that the detailed investigation on the radiation sources should be performed conservatively and accurately in the process of radiation shielding analysis for a BUC cask.

• Journal of the Korean Society of Radiology
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• v.9 no.5
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• pp.295-300
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• 2015

There are various type of particle accelerators such as Kyoungju 100-MeV proton beam accelerator in Korea. And Korea plans to build large particle accelerator such as heavy ion accelerator and 4th generation light source facility. The accelerated high energy particles of these facility produce 2nd neutron after nuclear reaction with target materials. And then these 2nd neutron activate structural materials and surrounding environment. Accordingly, it is important to consider the activation and shielding calculation on design of facility for safety operation. In this study, we tried to calculate and compare the neutron flux from the interaction $^{la}150$ beam with target material(Cu) according to thickness of iron and concrete shielding material by MCNPX 2.7 with nuclear library JENDL/HE 07and la150. To verify the properties of nuclear library, we compared computational results with experimental value. These results can be used for dose evaluation technology in planning of the shielding of large particle accelerator.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.41-50
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• 2006

In this paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm-container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by $\sim20$ tons.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.153-162
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• 2005

In this Paper, sets of engineering analyses were conducted to renew the overall dimensions and configurations of a disposal container proposed as a prototype in the previous study. Such efforts and calculation results can provide new design variables such as the inner basket array type and thickness of the outer shell and the lid & bottom of a spent nuclear fuel disposal container. These efforts include radiation shielding and nuclear criticality analyses to check to see whether the dimensions of the container proposed from the mechanical structural analyses can provide a nuclear safety or not. According to the results of the structural analysis of a PWR disposal container by varying the diameter of the container insert, the Maximum Von Mises stress from the 102 cm container meets the safety factor of 2.0 for both extreme and normal load conditions. This container also satisfies the nuclear criticality and radiation safety limits. This decrease in the diameter results in a weight loss of a container by ${\~}$20 tons.

• Journal of Radiation Protection and Research
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• v.10 no.1
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• pp.41-49
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• 1985

Neutron Streaming analysis in 1300 MWe pressurized water reactor cavity was performed. In this calculation, the discrete ordinates transport codes, ANISN and DOT 3.5, and the Monte Carlo code, TRIPOLI-02 were used with the coupling code, DOTTRI. In this study IBM 3033 type computer was used. The calculated neutron fluxes and dose rates were compared with the measured data in a 900MWe pressurized water reactor cavity to show a good agreement, although some deviations in the results for each energy group were noticed. These results will be applied in the radiation shielding design of high capacity nuclear power reactors and, to the means of radiation protection in case of the reactor maintenance and the access of the reactor cavity.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.391-402
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• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.171-179
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• 1980

This paper presentss flux-to-dose conversion factors for neutrons and gamma-rays based on the concept of the maximum absorbed dose. Neutron flux-to-does-rate conversion factors for energies from 2.5$\times$10$^{-8}$ to 20 MeV are presented while the conversion factors for gamma-rays are given in the energy range of 0.01 to 15MeV. Flux-to-does-rate conversion factors, which were calculated under the assumption that the radiation energy distribution has nonlinearity in phantom, are different from those values obtained by monoenergetic radiation. Especially, these values obtained here were determined for the cross section libray such as DLC-23, DLC-27, and DLC-31. The flux-to-dose-rate conversion factors obtained in this work are in a good agreement with the values presented by American National Standard Institute (ANSI) N666. These results are used to calculate the dose rate distribution of neutron and gamma-ray in any radiation fields, and will be useful for the radiation shielding analysis, radiation protection and radiation dosimetry concerned with problems of continuous energy distribution.