• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.949-959
• /
• 2021

The current study aims to explore the shielding properties of multi-component borate-based glass series. Seven glass-samples with composition of (80-y)H₃BO₃-10ZnO-10Na₂O-yBaO where (y = 0, 5, 10, 15, 20, 25 and 30 mol.%) were synthesized by melt-quench method. Various shielding features for photons, neutrons, and protons were determined for all prepared samples. XCOM, Phy-X program, and SRIM code were performed to determine and explain several shielding properties such as equivalent atomic number, exposure build-up factor, specific gamma-ray constants, effective removal cross-section (Σ_R), neutron scattering and absorption, Mass Stopping Power (MSP) and projected range. The energy ranges for photons and protons were 0.015-15 MeV and 0.01-10 MeV, respectively. The mass attenuation coefficient (μ/ρ) was also determined experimentally by utilizing two radioactive sources (¹⁶⁶Ho and ¹³⁷Cs). Consistent results were obtained between experimental and XCOM values in determining μ/ρ of the new glasses. The addition of BaO to the glass matrix led to enhance the μ/ρ and specific gamma-ray constants of glasses. Whereas the remarkable reductions in Σ_R, MSP, and projected range values were reported with increasing BaO concentrations. The acquired results nominate the use of these glasses in different radiation shielding purposes.

• Journal of radiological science and technology
• /
• v.39 no.1
• /
• pp.89-97
• /
• 2016

As previous studies to proceed with the evaluation of the radioactive at linear accelerator's shielding concrete wall. And the shielding wall was evaluated the characteristics for the incoming neutron. As a result, the shielding wall is the average amount of incoming neutrons 10 MV 4.63E-7%, 15 MV 9.69E-6%, showed the occurrence of 20 MV 2.18E-5%. The proportion of thermal neutrons of which are found to be approximately 18-33%. The neutron generation rate can be seen as a slight numerical order. However, in consideration of the linear accelerator operating time we can not ignore the effects of neutrons. Accordingly radioactive problem of the radiation shield wall of the treatment room will be this should be considered.

• Journal of the Korean Society of Radiology
• /
• v.15 no.2
• /
• pp.191-199
• /
• 2021

The purpose of this research is to analyze the radiation shielding effect of soft magnetic material to confirm the applicability to the military facilities. The soft magnetic material is known to be effective in shielding EMP. If this material is also effective in radiation shielding, it is expected that it has a lot of applicability in military protection. In particular, this material contains boron, so it will be effective in shielding neutrons. In this research, experiments were conducted using Cs-137 and Co-60 sources to check the gamma ray shielding effect. In addition, the Monte Carlo N-Particle(MCNP) modeling was applied to evaluate the gamma ray and neutron shielding effect of a military command tent. As a result, as the soft magnetic thickness increased, the shielding performance improved according the linear attenuation law of gamma ray and neutron. Therefore, this research verified that the application of soft magnetic material for military purposes in radiation shielding would be effective.

• Nuclear Engineering and Technology
• /
• v.40 no.4
• /
• pp.319-326
• /
• 2008

To transport process wastes efficiently from the Advanced Spent Fuel Conditioning Pyro-process Facility (ACPF) at the Korea Atomic Energy Research Institute (KAERI), a new hot cell cask has been designed based on an existing hot cell padirac transport cask, with not only a neutron absorber for improved shielding capability, but also a docking facility for an easy docking system. In the new hot cell cask, two kinds of materials have been considered as shielding materials, polyethylene and resin. To verify the transport compatibility of the waste and spent fuel for the ACPF, neutron and photon shielding calculations were performed using the MCNPX code. The source term was evaluated by the ORIGEN-ARP code system based on spent PWR fuel. From the calculation, it was found that the maximum surface dose rates of the hot cell cask with the two candidates were estimated within the limit (2 mSv/hr).

• Nuclear Engineering and Technology
• /
• v.51 no.1
• /
• pp.275-283
• /
• 2019

Rare isotope beam facilities require shielding data in early stage of their design. There is much less shielding data on neutrons from the reactions between heavy ion beams and matter than the data on neutrons produced by protons. The purpose of the present work is to produce and thus increase the amount of shielding data on neutrons generated by high-energy heavy ion beams based on the RAON in-flight fragment facility. Calculations were performed with the computational Monte Carlo codes PHITS and MCNPX. The secondary neutron source terms were evaluated at 550 MeV/u for Ca, Kr, and Sn and at 400 MeV/u for U ions on a graphite target. Source terms and attenuation lengths were obtained by fitting the ambient dose equivalent inside an ordinary concrete shield.

• Nuclear Engineering and Technology
• /
• v.52 no.7
• /
• pp.1355-1366
• /
• 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
• /
• v.25 no.3
• /
• pp.229-234
• /
• 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.

• Journal of the Korean Society of Radiology
• /
• v.13 no.1
• /
• pp.45-53
• /
• 2019

This study is a neutron activation for concrete that shields medical linear accelerator facilities. Comparison of general concrete and low activation concrete. The simulation method was simulated using MCNPX (Ver. 2.5.0) and FISPACT-2010, and the shielding ability for photon and neutron beams was calculated and neutron activation evaluation was carried out. As a result, the shielding capacity was 20 ~ 50 cm efficient in general concrete, and activate evaluation in low activation concrete was calculated to be low in radioactivity concrete, but all were estimated to not exceed their own allowable concentration in self - disposal. As a result of the comprehensive analysis, it is considered effective to use ordinary concrete.

• Applied Chemistry for Engineering
• /
• v.9 no.6
• /
• pp.884-888
• /
• 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 Korean Nuclear Society Conference
• /
• 1995.05a
• /
• pp.847-852
• /
• 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.