• Nuclear Engineering and Technology
• /
• v.36 no.2
• /
• pp.165-174
• /
• 2004

The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.244-250
• /
• 1996

Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.451-456
• /
• 1997

Temperature distribution measurements in the mockup apparatus of reactor vessel were performed to determine the effective thermal conductivity of porous media with different geometry and to obtain the experimental data for the heat transfer processes by natural convection occurring in the air duct. The temperature distributions at four separated sections with different arrangements of porous media have different slopes according to the geometrical configuration. From the measured temperature distribution, effective thermal conductivity have been derived using the least square fitting method. The test at air duct was performed to the high heat removal at 3.4kW/$m^2$ by the natural convection from the outer wall to the air. And also the temperature distributions in the air duct agree well with the 1/7th power-law turbulent temperature distribution. The obtained heat transfer data have been compared with the Shin's and Sieger's correlations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.3
• /
• pp.371-381
• /
• 2023

Evaluating the effectiveness of the radiation protection measures deployed at the Centralized Radioactive Waste Management Facility in Ghana is pivotal to guaranteeing the safety of personnel, public and the environment, thus the need for this study. Radiagem^TM 2000 was used in measuring the dose rate of the facility whilst the personal radiation exposure of the personnel from 2011 to 2022 was measured from the thermoluminescent dosimeter badges using Harshaw 6600 Plus Automated TLD Reader. The decay store containing scrap metals from dismantled disused sealed radioactive sources (DSRS), and low-level wastes measured the highest dose rate of 1.06 ± 0.92 µSv·h^-1. The range of the mean annual average personnel dose equivalent is 0.41-2.07 mSv. The annual effective doses are below the ICRP limit of 20 mSv. From the multivariate principal component analysis biplot, all the personal dose equivalent formed a cluster, and the cluster is mostly influenced by the radiological data from the outer wall surface of the facility where no DSRS are stored. The personal dose equivalents are not primarily due to the radiation exposures of staff during operations with DSRS at the facility but can be attributed to environmental radiation, thus the current radiation protection measures at the Facility can be deemed as effective.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.689-690
• /
• 2004

• Nuclear Engineering and Technology
• /
• v.37 no.6
• /
• pp.605-610
• /
• 2005

A method to estimate the relative sensitivity of a self-powered rhodium detector for an upcoming cycle is developed by combining the rhodium depletion data from a nuclear design with the site measurement data. This method can be used both by nuclear power plant designers and by site staffs of Korean standard nuclear power plants for determining which rhodium detectors should be replaced during overhauls.

• Earthquakes and Structures
• /
• v.12 no.6
• /
• pp.653-663
• /
• 2017

A seismic margin assessment evaluates how much margin exists for the system under beyond design basis earthquake events. Specifically, the seismic margin for the entire system is evaluated by utilizing a systems analysis based on the sub-system and component seismic fragility data. Each seismic fragility curve is obtained by using empirical, experimental, and/or numerical simulation data. The systems analysis is generally performed by employing a fault tree analysis. However, the current practice has clear limitations in that it cannot deal with the uncertainties of basic components and accommodate the newly observed data. Therefore, in this paper, we present a Bayesian-based seismic margin assessment that is conducted using seismic fragility data and fault tree analysis including Bayesian inference. This proposed approach is first applied to the pooltype nuclear research reactor system for the quantitative evaluation of the seismic margin. The results show that the applied approach can allow updating by considering the newly available data/information at any level of the fault tree, and can identify critical scenarios modified due to new information. Also, given the seismic hazard information, this approach is further extended to the real-time risk evaluation. Thus, the proposed approach can finally be expected to solve the fundamental restrictions of the current method.

• Journal of Radiation Protection and Research
• /
• v.26 no.3
• /
• pp.265-270
• /
• 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.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.137-138
• /
• 2005

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2012.10a
• /
• pp.127-128
• /
• 2012