• 한국건설순환자원학회논문집
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• 제4권3호
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• pp.269-275
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• 2016

본 연구에서는 붕규산 유리와 비정질 붕소강 섬유를 혼입한 콘크리트의 역학적 성능 및 중성자 차폐성능을 평가하였다. 잔골재를 붕규산 유리로 치환한 콘크리트와 비정질 붕소강 섬유를 보강한 콘크리트를 제조하여 압축강도, 정탄성계수, 압축인성, 휨강도, 휨인성, 중성자 차폐성능을 평가하였다. 실험결과, Plain 콘크리트와 대비하여 붕규산 유리를 혼입한 콘크리트의 경우 압축강도 및 휨강도는 저하되었으나, 중성자 차폐성능은 향상되는 것으로 나타났다. 비정질 붕소강 섬유를 혼입한 콘크리트의 경우 Plain 콘크리트와 대비하여 압축인성 및 휨인성이 증진되었으며 중성자 차폐성능도 향상되는 것으로 나타났다.

• 한국건축시공학회지
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• 제17권1호
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• pp.9-14
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• 2017

본 연구에서 비정질 붕소강 섬유를 혼입한 콘크리트의 역학적 성능 및 중성자 차폐성능을 평가하였다. 비정질 붕소강 섬유를 콘크리트 체적 대비 0.25%에서 1.0%까지 혼입하여 굳지 않은 콘크리트의 공기량과 슬럼프값, 경화된 콘크리트의 압축강도, 휨강도, 휨인성 및 중성자 차폐성능을 평가하였다. 실험결과, 비정질 붕소강 섬유의 혼입량이 증가할수록 콘크리트의 휨인성 및 중성자 차폐성능이 향상되는 것으로 나타났다. 이를 통해 비정질 붕소강 섬유의 혼입이 중성자 차폐성능 뿐만 아니라 역학적 성능을 효과적으로 개선시켜 줄 것이라고 기대된다.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• Journal of Radiation Protection and Research
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• 제22권2호
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• pp.77-83
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• 1997

사용 후 핵연료 수송용기 등에 사용되는 에폭시수지계 중성자 차폐재, KNS(Kaeri Neutron Shield)-101, KNS-102 및 KNS-103를 제조하였다. 기본물질은 에폭시수지이며, 첨가제로는 폴리프로필렌, 수산화알루미늄 및 탄화붕소이다. 이들 중성자 차폐재들은 유동성이 좋아 수송용기와 같은 복잡한 구조에 사용할 수 있다. 제조된 중성자 차폐재들을 가압경수로 사용 후 핵연료 28다발을 수송할 수 있는 수송용기에 적용하여 차폐능 평가를 수행하였다. 세가지 중성자 차폐재를 수송용기에 적용하여 ANISN 코드로 차폐능 평가를 수행한 결과 정상수송시 중성자 차폐재의 두께가 10 cm 이상 일때 수송용기 반경방향표면에서 최대 방사선량율은 $300{\mu}Sv/h$로 나타났으며, 수송용기 표면에서 100 cm 지점에서의 최대 방사선량율은 $97{\mu}Sv/h$로 나타났다. 이들은 모두 관련된 법규들에서 규정된 최대허용 방사선량율을 만족하는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1998-2004
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• 2019

The Multi-Physics (MP) instrument is one of 20 neutron spectrometers planned in the China Spallation Neutron Source (CSNS). This paper presents a shielding calculation for the MP instrument using Monte Carlo codes MCNPX and FLUKA. First, the neutrons that escape from the CSNS decoupled water moderator and are delivered to the beam line of the MP instrument are calculated to use as the source term of the shielding calculation. Then, to validate the calculation method based on multiple variance reduction techniques, a cross check between MCNPX and FLUKA codes is performed by comparing the calculation results of the dose rate distribution on a simplified beam line model. Finally, a complete geometry model of the MP instrument is set up, and the primary parameters for the shielding design are obtained according to the calculated dose rate map considering different worst-case scenarios.

• Nuclear Engineering and Technology
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• 제56권8호
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• pp.3307-3314
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• 2024

For traditional spent fuel shielding materials, due to physical and chemical defects and cost constraints, they have been unable to meet the needs. Therefore, this paper carries out the first discussion on the application and performance of bismuth in neutron shielding by establishing Monte Carlo simulation on the neutron flux model of shielded spent fuel. Firstly, functional fillers such as bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide are added to the matrices to compare the shielding rates of aluminum alloy matrix and silicone rubber matrix. The shielding rate of silicone rubber mixture is higher than aluminum alloy mixture, reaching more than 56%. The optimal addition proportion of bismuth oxide and lead oxide is 30%, and the neutron radiation protection efficiency reaches 60%. Then, the mass attenuation coefficients of bismuth oxide, lead oxide, boron oxide, gadolinium oxide and tungsten oxide in silicone rubber matrix are simulated with the change of functional fillers proportion and neutron energy. This simulation result shows that the mixture with functional fillers has good shielding performance for low energy neutrons, but poor shielding effect for high energy neutrons. Finally, in order to further evaluate the possibility of replacing lead oxide with bismuth oxide as shielding material, the half-value layers and various properties of bismuth oxide and lead oxide are compared. The results show that the shielding properties of bismuth oxide and lead oxide are basically the same, and the mechanical properties, heat resistance, radiation resistance and environmental protection of bismuth oxide are better than that of lead oxide. Therefore, in the case of neutron source strengths in the range of 0.01-6 MeV and secondary gamma rays produced below 2.5 MeV, bismuth can replace lead in neutron shielding applications.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2015년도 춘계 학술논문 발표대회
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• pp.160-161
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• 2015

Borosilicate glass can be used for improving neutron shielding of concrete. The well known expansion of borosilicate glass caused by expansion of mortar bar was can cause serious damage to the concrete. In this research, borosilicate glass was powdered to reduce the particle size similar to that of cement, and 20% cement replacement set was reduced expansion rate about 30%. But aggregate replacement set was damaged because of Alkali-Silica Reaction expansion.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.197-201
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• 1996

In the detail reaction-rate measurements in a critical assembly using the foil activation method, the measured activations of detector foils have inevitably errors caused by detector foil self-shielding effect. If neutron flux could be approximated to Westcott flux: i.e. well thermalized Maxwellian distribution, these activations of detector foil could be corrected to represent the unperturbated flux at any detected position in the cell with using Westcott option and reaction-rate option of the lattice code, WIMS-AECL. These calculated detector material self-shielding correction factors of the tested fuel, CANFLEX provided much information about neutron spectrum of test lattice cell as well as the correction factors themselves. The results could be verified by another lattice calculations.