• Journal of Radiation Protection and Research
• /
• v.26 no.3
• /
• pp.225-229
• /
• 2001

In Korea, an advanced reactor system of 330MWt power called SMART (System integrated Modular Advanced ReacTor) is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. A shielding design of the SMART core with 49 CEDM is established by a two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the SMART shielding designs. Three axial regions represent the SMART reactor assembly, each of which is modeled in the R-Z geometry. The BUGLE-96 library is used in the analyses, which consists of 47 neutron and 20 gamma energy groups. The results indicate that the maximum neutron fluence at the bottom of reactor vessel is $5.89 {\times} 10^{17}\;n/cm^2$ and that on the radial surface of reactor vessel is $4.49 {\times} 10^[16}\;n/cm^2$. These results meet the requirement, $1.0 {\times} 10^{20}\;n/cm^2$, in 10 CFR 50.61 and the integrity of SMART reactor vessel during the lifetime of the reactor is confirmed.

• Nuclear Engineering and Technology
• /
• v.4 no.1
• /
• pp.39-45
• /
• 1972

In neutron shielding, the scattering effect is equally important as the attenuations in shielding materials. In the present study, the scattered dose equivalent was measured using a Rem counter for water, paraffin, borated paraffin, ordinary and heavy concrete, lead, iron, and tissue equivalent material in three different angles: 45$^{\circ}$, 90$^{\circ}$, and 135$^{\circ}$, respectively. The measurements were performed for the neutron, having the energies of 0.5, 1, 2, 5, and 18 MeV, which are produced from the Van do Graaff accelerator. The scattered dose equivalent ratios were increased with increasing the thickness of scattering materials and saturated at a certain thickness although they were different from one to other materials under study. The ratios were large for lead and iron while they were small for the hydrogen containing materials such as water and paraffin etc.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3073-3084
• /
• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Nuclear Engineering and Technology
• /
• v.53 no.11
• /
• pp.3816-3823
• /
• 2021

The mechanical features evaluated theoretically using Makishima-Mackenzie's model for glasses xBaO-(50-x) PbO-50P₂O₅ where x = 0, 5, 10, 15, 20, 30, 40, and 50 mol%. Wherefore, the elastic characteristics; Young's, bulk, shear, and longitudinal modulus calculated. The obtained result showed an increase in the calculated values of elastic moduli with the replacement of the PbO by BaO contents. Moreover, the Poisson ratio, micro-hardness, and the softening temperature calculated for the investigated glasses. Besides, gamma and neutron shielding ability evaluated for the barium doped lead phosphate glasses. Monte Caro code (MCNP-5) and the Phy-X/PSD program applied to estimate the mass attenuation coefficient of the studied glasses. The decrease in the PbO ratio has a negative effect on the MAC. The highest MAC decreased from 65.896 cm²/g to 32.711 cm²/g at 0.015 MeV for BPP0 and BPP7, respectively. The calculated values of EBF and EABF showed that replacement of PbO with BaO contents in the studied BPP glasses helps to reduce the number of photons accumulated inside the studied BPP glasses.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.125-126
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.73-73
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.222.1-222
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.62-62
• /
• 2002

• Journal of Radiation Protection and Research
• /
• v.19 no.1
• /
• pp.13-22
• /
• 1994

The MCNP 4.2 code was used to calculate the thermal neutron flux distributions for $(n,\;{\gamma})$reaction in mainshell, annular plate, and subshell of the calandria of a CANDU 6 plant during operation. The thermal neutron flux distributions in calandria mainshell, annular plate, and subshell were in the range of $10^{11}{\sim}10^{13}\;neutrons/cm^2-sec$ which is somewhat higher than the previous estimates calculated by DOT 4.2 code. As an application to shielding analysis, photon dose rates outside the side and bottom shields were calculated. The resulting dose rates at the reactor accessible areas were below design target, $6 {\mu}Sv/h$. The methodology used in this study to evaluate the thermal neutron flux distribution for $(n,\;{\gamma})reaction$ can be applied to radiation shielding analysis of CANDU 6 type plants.

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