• Journal of Sensor Science and Technology
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• v.25 no.4
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• pp.264-270
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• 2016

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1036-1044
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• 2023

The combination of MCNP6 and the FISPACT codes was used to predict inventories of radioisotopes produced by neutron exposure of a sample in a research reactor. The detailed MCNP6 model of the Budapest Research Reactor and the specific irradiation geometry of the NAA channel was established, while realistic material cards were specified based on concentrations measured by PGAA and NAA, considering the precursor elements of all significant radioisotopes. The energy- and spatial distributions of the neutron field calculated by MCNP6 were transferred to FISPACT, and the resulting activities were validated against those measured using neutron-irradiated small and bulky targets. This approach is general enough to handle different target materials, shapes, and irradiation conditions. A general agreement within 10% has been achieved. Moreover, the method can also be made applicable to predict the activation properties of the near-vessel concrete of existing nuclear installations or assist in the optimal construction of new nuclear power plant units.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.549-555
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• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.178-184
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• 2020

The traditional methods for radiation shield design always only focus on either the structure or the components of the shields rather than both of them at the same time, which largely affects the shielding performance of the facilities, so in this paper, a novel method for designing the structure and components of shields simultaneously is put forward to enhance the shielding ability. The method is developed by using the genetic algorithm (GA) and the MCNP software. In the research, six types of shielding materials with different combinations of elements such as polyethylene (PE), lead (Pb) and Boron compounds are applied to the radiation shield design, and the performance of each material is analyzed and compared. Then two typical materials are selected based on the experiment result of the six samples, which are later verified by the Compact Accelerator Neutron Source (CANS) facility. By using this method, the optimal result can be reached rapidly, and since the design progress is semi-automatic for most procedures are completed by computer, the method saves time and improves accuracy.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1783-1790
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• 2023

Barium Bismuth Oxide Borate (BBOB) has been synthesized for the first time using solution combustion technique. SEM analysis reveal flower shape of the nanoparticles. The formation of the nanoparticles has been confirmed through XRD & FTIR studies which gives the physical and chemical structure of the novel material. The UV light absorption is observed in the range 200-300 nm. The present study highlights the radiation shielding ability of BBOB for different radiations like X/Gamma rays, Bremsstrauhlung and neutrons. The gamma shielding efficiency is comparable to that of lead in lower energy range and lesser than lead in the higher energy range. The bremsstrauhlung exposure constant is comparably larger for BBOB NPs than that of concrete and steel however it is lesser than that of lead. The beauty of BBOB nanoparticles lies in, high absorption of radiations and low emission of secondary radiations when compared to lead. In addition, the neutron shielding parameters like scattering length, absorption and scattering cross sections of BBOB are found to be much better than lead, steel and concrete. Thus, BBOB nanoparticles are highly efficient in absorbing X/Gamma rays, neutrons and bremsstrauhlung radiations.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.08a
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• pp.314-314
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• 2010

For the optimal design of a tokamak-type reactor, self-consistent determination of a radial build of reactor systems is important and the radial build has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor systems. In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the shield should provide sufficient protection for the superconducting TF coil and the shield plays a key role in determining the size of a reactor. To determine the radial build of a reactor, neutronic effects such as tritium breeding in the blanket, nuclear heating, and radiation damage to toroidal field (TF) coil has to be included in the systems analysis. In this work, the outboard blanket only is considered where tritium self-sufficiency is possible by using an inboard neutron reflector instead of breeding blanket. The reflecting shield should provide not only protection for the superconducting TF coil but also improved neutron economy for the tritium breeding in outboard blanket. Tungsten carbide, metal hydride such as titanium hydride and zirconium hydride can be used for improved shielding performance and thus smaller shield thickness. With the use of advanced technology in the shield, conceptual design of a compact superconducting LAR reactor with aspect ratio of less than 2 will be presented as a viable power plant.

• Journal of Radiation Protection and Research
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• v.30 no.2
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• pp.55-60
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• 2005

The evaluation of neutron flux distribution was performed for the ex-core detector design of SMART-P. DORT and MCNP code were used for the calculation of energy-dependent neutron flux distribution at 100% full power condition. Two code results show that maximum thermal flux appears at the $1^{st}$ water region in IST region and agree within 10% difference. In addition, another evaluation was performed code with assumptions that cote was composed of fission source and control rod without fuel assemblies. These assumptions make neutron count rate to be minimized. As a results, maximum thermal flux showed $6.99{\times}10^{-2}(n/cm^2-sec)$, when the strength of initial fission source was assumed as $1.0{\times}10^8(n/sec)$. The main reason of these results is due to the thermalization of fast neutrons in the water region and thermal flux is proportional to 80% of total neutron flux. Therefore, optimization of filler material of detector guide tube, position of installation and axial length of detector segments is necessary for the design of ex-core detector to enhance the neutron count rate and above results could be used in ex-core detector design as a fluence requirement.

• Proceedings of the Korean Vacuum Society Conference
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• 2011.02a
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• pp.397-397
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• 2011

In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the radial build of TF coil and the shield play a key role in determining the size of a reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with one-dimensional radiation transport code. Conceptual design study of a compact superconducting LAR tokamak reactor with aspect ratio less than 2.5 was conducted and the optimum radial build was identified. It is shown that the use of an improved shielding material and high temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at low aspect ratio, and that by using an inboard neutron reflector instead of breeding blanket, tritium self-sufficiency is possible with outboard blanket only and thus compact sized reactor is viable.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1887-1892
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• 2021

The effect of gadolinium and boron on the texture development and magnetic properties of the texture controlled 23Cr-10NiCu duplex stainless steels were studied to develop a high performance neutron and electromagnetic shielding material. The 23Cr-10NiCu base alloy is composed of 60% of austenite and 40% of ferrite, whereas, the 23Cr-10NiCu-0.5Gd-0.8B modified alloy is composed of 66% of austenite, 27% of ferrite and 7% of CrFeB intermetallic compounds. The gadolinium and boron addition to the 23Cr-10NiCu base alloy increased mechanical properties. Microstructure observation showed that the small addition of 0.5 wt% gadolinium and 0.8 wt% boron to the alloy retarded to form texture at the same hot rolling conditions, and improved the maximum magnetism, residual magnetism and coercive force about 3%, 122% and 120%, respectively.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2335-2347
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• 2023

As medical facilities are usually built at urban areas, special concrete aggregates and evaluation methods are needed to optimize the design of concrete walls by balancing density, thickness, material composition, cost, and other factors. Carbon treatment rooms require a high radiation shielding requirement, as the neutron yield from carbon therapy is much higher than the neutron yield of protons. In this case study, the maximum carbon energy is 430 MeV/u and the maximum current is 0.27 nA from a hybrid particle therapy system. Hospital or facility construction should consider this requirement to design a special heavy concrete. In this work, magnetite is adopted as the major aggregate. Density is determined mainly by the major aggregate content of magnetite, and a heavy concrete test block was constructed for structural tests. The compressive strength is 35.7 MPa. The density ranges from 3.65 g/cm³ to 4.14 g/cm³, and the iron mass content ranges from 53.78% to 60.38% from the 12 cored sample measurements. It was found that there is a linear relationship between density and iron content, and mixing impurities should be the major reason leading to the nonuniform element and density distribution. The effect of this nonuniformity on radiation shielding properties for a carbon treatment room is investigated by three groups of Monte Carlo simulations. Higher density dominates to reduce shielding thickness. However, a higher content of high-Z elements will weaken the shielding strength, especially at a lower dose rate threshold and vice versa. The weakened side effect of a high iron content on the shielding property is obvious at 2.5 µSv=h. Therefore, we should not blindly pursue high Z content in engineering. If the thickness is constrained to 2 m, then the density can be reduced to 3.3 g/cm³, which will save cost by reducing the magnetite composition with 50.44% iron content. If a higher density of 3.9 g/cm³ with 57.65% iron content is selected for construction, then the thickness of the wall can be reduced to 174.2 cm, which will save space for equipment installation.