• Journal of Radiation Industry
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• v.18 no.1
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• pp.15-21
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• 2024

The Non-Proliferation Treaty (NPT) is the basis of global efforts to prevent the spread of nuclear weapons. In Republic of Korea, safety measures are integrated with NPT approval through agreements with the International Atomic Energy Agency (IAEA) and the Safeguards Agreement. In contrast, Democratic People's Republic of Korea (DPRK), initially an NPT member, withdrew, refusing IAEA nuclear inspections. This inhibits the precise management of DPRK's nuclear facilities and limits access to related information. The Korean Peninsula, politically divided, sees DPRK in control of nuclear weapons. Although the IAEA periodically evaluates DPRK's nuclear facilities, there's a research gap in contamination and site management with nuclear activities. Recognizing the presence or absence of such activities is crucial for peaceful nuclear endeavors. This proposal suggests the number and locations for environmental sample collection using the Visual Sample Plan (VSP) software for nuclear activity analysis. VSP software is sample collection locations and quantities through statistical tests on collected data, ensuring reliability for decision-making. The proposal identifies sites and facilities for nuclear activity analysis based on IAEA safety reports, utilizing the software's embedded methods. Suggested sampling locations for undisclosed nuclear activities employ VSP's embedded techniques, including 'Show that at least some high % of the sampling area is acceptable' to confirm contamination and 'Estimate the Mean' to evaluate the average contamination level.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.403-408
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• 1996

The subchannel analysis was performed to verify the behavior of hot channel characteristics and obtain the information to support the core thermal-hydraulic behavior at post-trip steam line break with low flow condition. During this postulated accident, buoyancy-induced cross flow occurs, and the coupled nuclear and thermal-hydraulic interactions become important. The code predictions with TORC are in good agreement with the test data. Under such conditions, the mass flow increase in the hot channel by buoyancy-induced cross flow depends on the parameter $GR^{*}\;/\;Re^2$, and buoyancy effect becomes more noticeable as $GR^{*}\;/\;Re^2$ increases.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.683-690
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• 1996

The MIDLOOP code has been developed for the evaluation of RES pressurization transients initiated from a loss-of-Residual Heat Removal System (RHRS) during mid-loop operation after reactor shutdown. It provides a fast running and realistic tool for studying parametrically the response of important plant parameters such as pressure, temperature, and level to various plant combinations of the primary side vent, makeup, and leakage procedures and the steam generator (SG) conditions. The code consists of ten nodes representing the primary and secondary sides of a nuclear power plant and can analyze the effect of air on the primary system pressurization and primary to secondary heat transfer. The analysis results of the MIDLOOP code are in good agreement with the ROSA-IV/LSTF experiment without opening in the RCS.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.407-415
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• 2015

A developing post-dryout region is characterized by significant heat transfer enhancements compared with the fully developed post-dryout region. The heat transfer enhancements are mainly due to upstream disturbance and entrained droplets in the region immediately downstream of the critical heat flux location. In this paper, an improved heat transfer correlation is developed for the developing post-dryout regions in vertical tubes over a wide range of flow conditions. The correlation represents a correction factor for the fully developed film-boiling look-up table to be applied to the developing post-dryout region. The new correlation significantly improves the heat transfer prediction in the developing post-dryout regions and provides very good agreement with the experimental data.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.459-464
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• 2001

For major nuclear power plant components periodic inspections and integrity assessments are needed for the safety. But many flaws are undetectable due to sampling inspection. Probabilistic integrity assessment is applied to take into consideration of uncertainty and variance of input parameters arise due to material properties, applied load and undetectable flaws. This paper describes a Probabilistic Fracture Mechanics(PFM) analysis based on Monte Carlo(MC) algorithms. Taking important parameters as probabilistic variables such as fracture toughness, crack growth rate and flaw shape, failure probability of major nuclear power plant components is archived as a results of MC simulation. For the verification of these analysis, a comparison study of the PFM analysis using other commercial code, mathematical method is carried out and a good agreement was observed between those results.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.533-544
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• 2002

The neutron induced nuclear data for Mo-95, Tc-99, Ru-101 and Rh-103 was calculated and evaluated in the fast energy region. The energy dependent optical model potential parameters were extracted based on the recent experimental data and applied up to 20 MeV. The s-wave strength function was calculated from the parameters. Spherical optical model, statistical model in equilibrium energy, multistep direct and multistep compound model in pre-equilibrium energy and direct capture model were used in the calculation. The theoretically calculated cross sections were compared with the experimental data and the evaluated files The model- calculated total and capture cross sections were in good agreement with the reference experimental data. The direct capture contribution improved the capture cross sections in pre- equilibrium region. The evaluated cross section results were compiled to ENDF-6 format and will improve the ENDF/B-Vl.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.17-33
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• 2000

A mechanistic model based on wall-attached bubble coalescence, previously developed by the authors, was extended to predict a vow high critical heat flux (CHF）in highly subcooled flow boiling, especially for high mass flux and small tube diameter conditions. In order to take into account the enhanced condensation due to high subcooling and high mass velocity in small diameter tubes, a mechanistic approach was adopted to evaluate the non-equilibrium flow quality and void fraction in the subcooled water flow boiling, with preserving the structure of the previous CHF model. Comparison of the model predictions against highly subcooled water CHF data showed relatively good agreement over a wide range of parameters. The significance of the proposed CHF model lies in its generality in applying over the entire subcooled flow boiling regime including the operating conditions of fission and fusion reactors.

• Nuclear Engineering and Technology
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• v.35 no.1
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• pp.35-44
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• 2003

The neutron induced nuclear data for Eu-153, Gd-155 and Cd-157 are calculated and evaluated in the high energy region. The evaluation procedure for deformed nuclei is setup by using Ecis-Empire codes. The energy dependent optical model potential parameters are searched based on the recent experimental data and applied up to 20 MeV. Optical model, full featured Hauser-Feshbach model and multistep direct and multistep compound model are used in the calculation. The direct-semidirect capture model and the direct coupled-channels contribution to discrete levels are introduced to improve the capture and inelastic scattering cross sections. The theoretically calculated cross sections are compared with the experimental data and the evaluated files. The model-calculated total and capture cross sections are in good agreement with the reference experimental data. The evaluated cross section results are compiled to ENDF-6 format and are expected to improve the ENDF/B-Vl.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1252-1263
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• 2016

This paper presents a numerical method to simulate ductile tearing in cracked components under high strain rates using finite element damage analysis. The strain rate dependence on tensile properties and multiaxial fracture strain is characterized by the model developed by Johnson and Cook. The damage model is then defined based on the ductility exhaustion concept using the strain rate dependent multiaxial fracture strain concept. The proposed model is applied to simulate previously published three cracked pipe bending test results under two different test speed conditions. Simulated results show overall good agreement with experimental results.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1036-1046
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• 2016

We propose a primary water stress corrosion cracking (PWSCC) initiation model of Alloy 600 that considers the stress triaxiality factor to apply to finite element analysis. We investigated the correlation between stress triaxiality effects and PWSCC growth behavior in cold-worked Alloy 600 stream generator tubes, and identified an additional stress triaxiality factor that can be added to Garud's PWSCC initiation model. By applying the proposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growth simulations based on the macroscopic phenomenological damage mechanics approach were carried out on the PWSCC growth tests of various cold-worked Alloy 600 steam generator tubes and compact tension specimens. As a result, PWSCC growth behavior results from the finite element prediction are in good agreement with the experimental results.