• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.265-270
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• 2001

Experimental data have been obtained on the neutron elastic scattering cross sections for 55, 65 and 75 MeV neutrons, and non-elastic scattering cross sections for 40 to 80 MeV neutrons using the $^7Li(p,n)$ neutron source at TIARA of Japan Atomic Energy Research Institute and the TOF method. Data were obtained for C, Si, Fe, Zr, and Pb of natural elements. Elastic scattering data were obtained for 25 laboratory angles between 2.6 and 53.0 that clarified the angular distributions and angle integrated values. The data obtained were compared favorably with recent LA150 data library.

• 한국산업정보학회:학술대회논문집
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• 한국산업정보학회 1999년도 춘계학술대회 발표논문집
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• pp.217-222
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• 1999

본 연구에서는 토목공사에서 필요한 성토시공관리용 방사성 동위원소 이용계기의 측정회로를 개발하고자 하였다. 제작된 계기는 감마선과 열중성자 검출회로, 고전압 공급장치 그리고 마이크로프로세서 등으로 구성하였다. 성토의 밀도측정에 충분한 계측수를 얻기 위하여 감마선 검출 5회로, 열중성자 검출 2회로로 구성하였으며, 또한, 모든 회로는 자연 방사선과 잡음에 의한 영향을 최소화하기 위하여 정전차폐하였으며, 계수관에 인가하는 고전압의 리플 진폭과 주파수를 고려하여 펄스계수시에 리플 성분에 의한 펄스수는 제거하였다.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.360-365
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• 2020

A technique for the reduction of pulse pile-up effect in digital pulse-shape discrimination (PSD) of neutrons and gamma-rays with organic scintillation detectors is presented. The technique is based on an electronic reduction of the effective decay-time constant of scintillation pulses while retaining the PSD information of the pulses. The experimental results obtained with a NE213 liquid scintillation detector in a mixed radiation field of neutrons and gamma-rays are presented, demonstrating a figure of merit (FOM) of 1.20 ± 0.05 with an energy threshold of 350 keVee (electron equivalent energy) when the effective length of the pulses is reduced to 50 ns.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.75-82
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• 2011

The neutron multiplication factor in spent fuel storage racks, in which a stainless steel plate encloses a fuel assembly, was evaluated according to the variation of distance between the fuel assembly and stainless steel plate, as well as the pitch. The stainless steel plate position with the lowest multiplication factor on each pitch consistently appeared as 6mm or 9mm away from the outmost surface of the fuel assembly. Because the stainless steel plate has a thermal neutron absorption cross section, its ability to absorb neutrons can work best only if it is installed at the position where thermal neutrons can be gathered most easily. Therefore, the stainless steel plate position should not be too close or too far away from the fuel assembly, but it should be kept a pertinent distance from the fuel assembly.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2018년도 추계 학술논문 발표대회
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• pp.114-115
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• 2018

To develop an effective shielding material for spent fuel that emits fast neutrons is necessary. In this study, thermal neutron and fast neutron shielding performance of polyethylene coated boron carbide-incorporated cement paste was quantitatively analyzed by Monte Carlo N-Particle transport code (MCNP) simulations. As the results of the simulations, fast neutrons were effectively shielded through large quantity of hydrogen and boron elements in polyethylene and boron carbide.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1199-1210
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• 2017

A new in-line method for sampling neutrons emitted in (${\alpha}$,n) reactions based on alpha particle source information has been developed for continuous-energy Monte Carlo simulations. The new method uses a continuous-slowing-down model coupled with (${\alpha}$,n) cross section data to precompute the expected neutron yield over the alpha particle lifetime. This eliminates the complexity and computational cost associated with explicit charged particle transport. When combined with an integrated alpha particle decay source sampling capability, the proposed method provides an efficient and accurate method for sampling (${\alpha}$,n) neutrons based solely on nuclide inventories in the problem, with no additional user input required. Results from several example calculations show that the proposed method reproduces the (${\alpha}$,n) neutron yields and energy spectra from reference experiments and calculations.

• Nuclear Engineering and Technology
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• 제4권2호
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• pp.102-108
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• 1972

등방성이고 단면적이 상수인 경우의 지발 중성자를 가진 시간 종속 중성자 수송 방정식이 해석적으로 풀어지고 있다. 두 개로 구분된 시간 영역에 있어서의 방정식이 점근적 방법에 의하여 원래의 수송 방정식으로부터 얻어 지고 있다. 각 시간 영역에 있어서의 근사해는 중성자 속도의 역수 정도로 시간에 있어서 균일하게 유용하다는 것이 보여 지고 있다.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4067-4071
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• 2021

The ¹³C (α,n)¹⁶O reaction cross-section is important data for nuclear physics, astrophysical, and neutrino physics experiments, however, they exhibit uncertainties due to the discrepancies in the experimental data. In this study, using the Nedis-2m program code, the energy spectrum of α-induced neutrons in a thin carbon target was calculated and the corresponding reaction cross-section was refined in the alpha particle energy range of 5-8 MeV. The results were used to calculate the intensity and energy spectrum of background neutrons produced in the liquid scintillator of KamLAND. The results will be useful in a variety of astrophysical and neutrino experiments especially those based on LS or Gd-LS detectors.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.943-948
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• 2019

In this study, investigation of spent fuel was performed using six group delayed neutron parameters. Three used fuels (F1, F2, and F11) which are burnt over the years in the core of Missouri University of Science and Technology Reactor (MSTR), were investigated. F16 fresh fuel was used as plutonium free fuel element and compared with irradiated used fuels to develop burnup and Pu discrimination method. The fast fission factor of the MSTR was calculated to be 1.071 which was used for burnup calculations. Burnup values of F2 and F11 fuel elements were estimated to be 1.98 g and 2.7 g, respectively. $^{239}Pu$ conversion was calculated to be 0.36 g and 0.50 g for F2 and F11 elements, respectively.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1723-1729
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• 2021

Energy and directional distribution of neutrons at the Beloyarsk NPP workplaces is a subject of this study. Measurements of H^*(10) rate and neutron energy distribution were taken at 8 workplaces, which can be divided into three categories: work with spent or fresh nuclear fuel, work with radionuclide neutron sources, work at the rooms adjusted to reactors. The Hp(10) measurements were performed only at 6 out of 8 locations, due to the fact that long term placing of an effective neutron moderator in fresh nuclear fuel storage facility is forbidden. As a result of the research energy and direction distribution of the neutron fields at 8 locations of the Beloyarsk NPP workplaces was obtained. To estimate the accuracy of the H^*(10) rate and Hp (10) measurements the reference values of dose equivalents were calculated using energy and directional distribution. To take into account the difference between the reference values and the measured results site-specific correction factors were calculated.