• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.611-622
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• 2012

"Scaling" plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a "roadmap" to scaling. Key elements are the "scaling-pyramid", related "scaling bridges" and a logical path across scaling achievements (which constitute the "scaling puzzle"). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the "key-to-scaling", in the licensing process of a nuclear power plant. The proposed "road map to scaling" aims at solving the "scaling puzzle", by introducing a unified approach to the problem.

• Journal of the Korean Society of Safety
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• v.20 no.3 s.71
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• pp.84-90
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• 2005

Thermal decomposition of the copolymer of ${\alpha}$-Methylstyrene(AMS) with Acrylonitrile(AN) was investigated. The copolymer was synthesized in a continuous stirred tank reactor(CSTR) at $80^{\circ}C$ using toluene and benzoyl peroxide(BPO) as solvent and initiator, respectively. The reactor volume was 0.3 liters and residence time was 3 hours. The activation energy of thermal decomposition was in the ranges of $34{\sim}54$ kcal/mol for AMS with AN copolymer. The thermogravimetric trace curves were well agreed with the theoretical calculation.

• Geomechanics and Engineering
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• v.21 no.4
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• pp.379-390
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• 2020

The behavior of a piled raft system in multi-layered soil subjected to vertical loading has been studied numerically using 3D finite element analysis. Initially, the 3D finite element model has been validated by analytically simulating the field experiments conducted on vertically loaded instrumented piled raft. Subsequently, a comprehensive parametric study has been conducted to assess the performance of a combined piled raft system in terms of optimum pile spacing and settlement of raft and piles, in multi-layered soil stratum subjected to vertical loading. It has been found that a combined pile raft system can significantly reduce the total settlement as well as the differential settlement of the raft in comparison to the raft alone. Two different arrangements below the piled raft with the same pile numbers show a significant amount of increase of load transfer of piled raft system, which is in line with the load transfer mechanism of a piled raft. A methodology for the factor of safety assessment of a combined pile raft foundation has been presented to improve the performance of piled raft based on its serviceability requirements. The findings of this study could be used as guidelines for achieving economical design for combined piled raft systems.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.388-402
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• 2004

A phenomena identification and ranking table(PIRT) was developed for a main steam line break (MSLB) event for the Advanced Power Reactor-1400 (APR-1400). The selectee event was a double-ended steam line break at full power, with the reactor coolant pump running. The developmental panel selected the fuel performance as the primary safety criterion during the ranking process. The plant design data, the results of the APR-1400 safety analysis, and the results of an additional best-estimate analysis by the MARS computer code were used in the development of the PIRT. The period of the transient was composed of three phases: pre-trip, rapid cool-down, and safety injection. Based on the relative importance to the primary evaluation criterion, the ranking of each system, component, and phenomenon/process was performed for each time phase. Finally, the knowledge-level for each important process for certain components was ranked in terms of existing knowledge. The PIRT can be used as a guide for planning cost-effective experimental programs and for code development efforts, especially for the quantification of those processes and/or phenomena that are highly important, but not well understood.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• Environmental Engineering Research
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• v.24 no.1
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• pp.107-116
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• 2019

Nitrogen removal, nitrous oxide ($N_2O$) emission and microbial community in sequencing batch and continuous-flow intermittent aeration processes were investigated. Two sequencing batch reactors (SBRs) and two continuous-flow multiple anoxic and aerobic reactors (CMRs) were operated under high dissolved oxygen (DO) (SBR-H and CMR-H) and low DO (SBR-L and CMR-L) concentrations, respectively. Nitrogen removal was enhanced under CMR and low DO conditions (CMR-L). The highest total inorganic nitrogen removal efficiency of 91.5% was achieved. Higher nitrifying and denitrifying activities in SBRs were observed. CMRs possessed higher $N_2O$ emission factors during nitrification in the presence of organics, with the highest $N_2O$ emission factor of 60.7% in CMR-L. SBR and low DO conditions promoted $N_2O$ emission during denitrification. CMR systems had higher microbial diversity. Candidatus Accumulibacter, Nitrosomonadaceae and putative denitrifiers ($N_2O$ reducers and producers) were responsible for $N_2O$ emission.

• Journal of the Korean Society of Safety
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• v.36 no.2
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.1954-1961
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• 2022

Lead-based fast reactor (LFR) has become one of the most promising reactors for Generation IV nuclear systems. A developing trend of LFR is high efficiency, along with operation temperatures up to 800 ℃ or even higher. One of key issues in the high-efficiency LFR is corrosion of cladding materials with lead at high temperatures. In this study, corrosion behavior of some refractory metals (Nb, Nb521, and Mo-0.5La) was investigated in static lead at 1000 ℃ for 1000 h. The results showed that Nb and Nb521 exhibited an intense dissolution corrosion with obvious lead penetration after corrosion, and lead penetration extended along the grain boundaries of the specimens. Furthermore, Nb521 showed a better corrosion resistance than that of Nb as a result of the elements of W and Mo included in Nb521. Mo-0.5La showed much better corrosion resistance than that of Nb and Nb521, and no lead penetration could be observed. However, an etched morphology appeared on the surface of Mo-0.5La, indicating the occurrence of corrosion to a certain degree. The results indicate that Mo-0.5La is compatible with lead up to 1000 ℃. While Nb and Nb alloys might be not compatible with lead for high-efficiency LFR at such high temperatures.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3324-3335
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• 2022

Several countries show interest in the Generation-IV power reactor innovative small module (PRISM), including: Canada, Japan, Korea, Saudi Arabia and the United Kingdom. Generation IV International Forum (GIF) has recommended the utilizing of probabilistic safety assessment (PSA) in evaluating the safety of Generation-IV reactors. This paper reviews the PSA performed for PRISM using SAPHIRE 7.27 code. This work shows that the core damage frequency (CDF) of PRISM for a single module is estimated by 8.5E-8/year which is lower than the Generation-IV target that is 1E-6 core damage per year. The social risk of PRISM (likelihood of latent cancer fatality) with evacuation is estimated by 9.0E-12/year which is much lower than the basic safety objective (BSO) that is 1E-7/year. The social risk without evacuation is estimated by 1.2E- 11/year which is also much lower than the BSO. For the individual risk (likelihood of prompt fatality), it is concluded that it can be considered negligible with evacuation (1.0E-13/year). Assuming no evacuation, the individual risk is 2.7E-10/year which is again much lower than the BSO. In comparison with other PSAs performed for similar sodium fast reactors (SFRs), it shows that PRISM concept has the lowest CDF.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2844-2853
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• 2023

A sophisticated Bayesian multilevel model for estimating group bias was developed to improve the utility of the ASTM E900-15 embrittlement trend curve (ETC) to assess the conditions of nuclear power plants (NPPs). For multilevel model development, the Baseline 22 surveillance dataset was basically classified into groups based on the NPP name, product form, and notch orientation. By including the notch direction in the grouping criteria, the developed model could account for TTS differences among NPP groups with different notch orientations, which have not been considered in previous ETCs. The parameters of the multilevel model and biases of the NPP groups were calculated using the Markov Chain Monte Carlo method. As the number of data points within a group increased, the group bias approached the mean residual, resulting in reduced credible intervals of the mean, and vice versa. Even when the number of surveillance test data points was less than three, the multilevel model could estimate appropriate biases without overfitting. The model also allowed for a quantitative estimate of the changes in the bias and prediction interval that occurred as a result of adding more surveillance test data. The biases estimated through the multilevel model significantly improved the performance of E900-15.