• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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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.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1444-1450
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• 2019

The fast-neutron shielding behaviour (FNSB) of two clay-materials (Ball clay and Kaolin)of Southwestern Nigeria ($7.49^{\circ}N$, $4.55^{\circ}E$) have been investigated using effective removal cross section, ${\Sigma}_R(cm^{-1})$, mass removal cross section, ${\Sigma}_{R/{\rho}}(cm^2g^{-1})$ and Mean free path, ${\lambda}$ (cm). These parameters decide neutron shielding behaviour of any material. A computer program - WinNC-Toolkit has been used for computation of these parameters. The toolkit evaluates these parameters by using elemental compositions and densities of samples. The proficiency of WinNC-Toolkit code was probe by using MCNPX and GEANT4 to model fast neutron transmission of the samples under narrow beam geometry, intending to represent the actual experimental setup. Direct calculation of effective removal cross section ($cm^{-1}$) of the samples was also carried out. The results from each of the methods for each types of the studied clay-materials (Ball clay and Kaolin) shows similar trend. The trend might be the fingerprint of water content retained in each of the samples being baked at different temperature. The compositions of each sample have been obtained by Particle-Induced X-ray Emission (PIXE) technique (Tandem Pelletron Accelerator: 1.7 MV, Model 5SDH). The FNSB of the selected clay-materials have been compared with standard concrete. The cognizance of various factors such as availability, thermo-chemical stability and water retaining ability by the clay-samples can be analyzed for efficacy of the material for their FNSB.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1355-1366
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• 2020

A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed benchmarks describe satisfactorily the energy resolution of the photon detector and the geometry of the spherical samples with inner beam tube, tritium target and cooling water circuit, but lack information regarding the detector geometry and the distances of shields and collimators relatively to the neutron source and the detector. Calculations are therefore conducted for a sphere model only. A preliminary verification of MCS neutron-photon calculations against MCNP6.2 is first conducted, then the impact of modelling the inner beam tube, tritium target and cooling water circuit is assessed. Finally, a comparison of calculated results with the libraries ENDF/B-VII.1 and ENDF/B-VIII.0 against the measurements is conducted and shows reasonable agreement. The MCS and MCNP inputs used for the interpretation are available as supplementary material of this article.

• Journal of Sensor Science and Technology
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• v.25 no.3
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• pp.229-234
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• 2016

Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the vanadium (V) SPND has been being developed to be used in OPR1000 nuclear power plants. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina insulator with a cylindrical geometry. An MCNP code was used to simulate some factors (neutron self-shielding factor and beta escape probability from the emitter) and space charge effect of an insulator necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND and contribute to the development of TMI (Top-mount In-core Instrumentation) which might be used in the SMART and SMR.

• Applied Chemistry for Engineering
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• v.7 no.3
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• pp.597-604
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• 1996

Resin-type neutron shielding materials, KNS-115A, 115B and 115C have been fabricated to be used for spent fuel shipping cask. The base material is epoxy resin, and polypropylene, aluminium hydroxide and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to apply this resin shield to complicated geometric shapes such as shipping cask. Several measurements were made for the shielding materials to evaluate the shielding property, combustion characteristics, fire resistance, thermal and mechanical properties. The neutron shielding ability of the shielding materials is estimated to be better than that of foreign's shielding material, NS-4-FR, due to higher hydrogen atomic density. Other properties of the shielding materials are as follows: Onset temperatures; $267{\sim}270^{\circ}C$, thermal conductivities; $0.62{\sim}0.72W/m{\cdot}K$, combustion characteristics; <$800^{\circ}C$, ATB(average time of burning); <5sec, AEB(average extent of burning) ; <5mm, tensile strengths; $2.3{\sim}3.0kg/mm^2$, compressive strengths; $5.3{\sim}13.3kg/mm^2$, flexural strengths; $4.4{\sim}5.4kg/mm^2$.

• Korean Journal of Materials Research
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• v.8 no.5
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• pp.457-463
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• 1998

New neutron shielding materials, KNS-201, KNS-301 and KNS-601 have been fabricated to be used for radioactive material shipping and storage cask. The base materials are a modified and a hydrogenated bisphenol- A type and novolac type epoxy resin, and aluminium hydroxide and boron carbide are added. These shielding materials offer good fluidity at processing, which makes it possible to form this resin shield into complicated geometric shapes such as radioactive material shipping and storage cask. Several measurements were made for the shielding materials to evaluate the thermal and mechanical properties and radiation resistance. The properties of the shielding materials are as follows: onset temperatures 2S7~28$0^{\circ}C$, thermal conductivities 0.9S~1.14W/m. K, thermal expansion coefficients 0.77~1.26x$10_{-6}{\circ}C_{-1}$, combustion characteristics < 80$0^{\circ}C$, ATB(average time of burning) < 5sec, AEB(average extent of burning) < 5mm, tensile strengths 2.5~3.2kg/$\textrm{mm}^2$, compressive strengths 13.2~1S.2kg/$\textrm{mm}^2$, flexural strengths 5.2 -6.4kg/$\textrm{mm}^2$. In general, the concerned properties of KNS-201, KNS-301 and KNS-601 were revealed to be better than those of NS-4- FR. foreign neutron shielding material. It is also observed that the radiation resistance of KNS- 601 was better than those of KNS-201 and KNS-301.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.921-928
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• 2013

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

• Applied Chemistry for Engineering
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• v.9 no.6
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• pp.884-888
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• 1998

Effects of heating time under high temperature on the thermal and mechanical properties of neutron shielding materials based on modified (KNS-102), hydrogenated(KNS-106) bisphenol-A type epoxy resin and phenol-novolac(KNS-611) type epoxy resin for radioactive material shipping casks have been investigated. At early stages, the initial decomposition temperatures of the shielding materials of KNS-102, KNS-106 and KNS-611 increased with the heating time under high temperature, but it was rarely affected by the heating time in the later stages. In addition, the thermal conductivities of KNS-102 and KNS-106 decreased with heating time, but that of KNS-611 increased with the heating time. On the contrary, the thermal expansion coefficients of neutron shielding materials decreased with increase of heating time. At the high temperature, the tensile strength and flexural strength of the shielding materials of KNS-102 and KNS-611 increased with heating time, but those of KNS-106 decreased with increase of heating time. And the heating time under high temperature on the neutron shielding materials did not show measurable loss of weight and hydrogen content.

• Proceedings of the Materials Research Society of Korea Conference
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• 2011.05a
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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.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.470-477
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• 2018

With a highly functional methyl vinyl silicone rubber (VMQ) matrix and filler materials of $B_4C$, PbO, and benzophenone (BP) and through powder surface modification, silicone rubber mixing, and vulcanized molding, a flexible radiation shielding and resistant composite was prepared in the study. The dispersion property of the powder in the matrix filler was improved by powder surface modification. BP was added into the matrix to enhance the radiation resistance performance of the composites. After irradiation, the tensile strength, elongation, and tear strength of the composites decreased, while the Shore hardness of the composites and the crosslinking density of the VMQ matrix increased. Moreover, the composites with BP showed better mechanical properties and smaller crosslinking density than those without BP after irradiation. The initial degradation temperatures of the composites containing BP before and after irradiation were $323.6^{\circ}C$ and $335.3^{\circ}C$, respectively. The transmission of neutrons for a 2-mm thick sample was only 0.12 for an Am-Be neutron source. The transmission of ${\gamma}$-rays with energies of 0.662, 1.173, and 1.332 MeV for 2-cm thick samples were 0.7, 0.782, and 0.795, respectively.