• Journal of Environmental Science International
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• v.10 no.1
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• pp.65-71
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• 2001

HRP(high rate pond) which had kept the manufactured clay of 3cm-thickness as benthic clay in reactor and the 6 flat-blade turbine as impeller for agitation was named HRASP(high rate algae stabilization pond). And the experiment for treatment of artificial synthesis wastewater containing COD :300mg/$\ell$, NH$_3$-N : 300mg/$\ell$, T-P : 9mg/$\ell$ as nutrients was been performed successfully. This reactor was been operated under conditions : 24hrs.-irradiation and water temperature, $25^{\circ}C$ and pH 7 and agitation velocity, 15, 30, 45rpm and the effect of agitation velocity on algal bioaccumulation of nutrients was been studied with view point of fluid dynamics. The next followings could be obtained as results. 1. The agitation with a turbine impeller blade in HRASP makes clay particle indicate superior suspension effect by means of forming of excellent curl/shear flow in reactor. 2. The excessive suspension of clay particle which is created at 45rpm as rotation velocity of impeller blade of turbine disturbs the light penetration and algal photosynthesis reaction. 3. Efficiencies for removal of nutrients come out as COD : 93.9%~94.3%, ($NH_3-N + NO_3-N$) : 81.9%~99.0%, T-P : 46.8%~53.6%. 4. Kuo values of $K_1$for algal growth come out seperately as 15rpm : $1.876{\times}10^{-2}, 30rpm : 4.618{\times}10^{-3}$. 5. Kuo values of $K_2$for removal of N, P come out seperately as 15rpm : $8.403{\times}10^{-1}$ and $1.397{\times}10^{-1}$, 30rpm : $4.823{\times}10^{-1} and 2.052{\times}10^{-1}$. 6. It can be guessed easily that the excessive agitation can inhibit the algal and bacterial symbiotic reaction if it is considered that micro organism\` sense to preservation of life is relied on natural function of metabolism. Therefore the studies for this matter should be followed continuously.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.26-32
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• 2015

During the course of a severe accident in a light water nuclear reactor, large amounts of hydrogen can be generated and released into the containment during reactor core degradation. Additional burnable gases [hydrogen ($H_2$) and carbon monoxide (CO)] may be released into the containment in the corium/concrete interaction. This could subsequently raise a combustion hazard. As the Fukushima accidents revealed, hydrogen combustion can cause high pressure spikes that could challenge the reactor buildings and lead to failure of the surrounding buildings. To prevent the gas explosion hazard, most mitigation strategies adopted by European countries are based on the implementation of passive autocatalytic recombiners (PARs). Studies of representative accident sequences indicate that, despite the installation of PARs, it is difficult to prevent at all times and locations, the formation of a combustible mixture that potentially leads to local flame acceleration. Complementary research and development (R&D) projects were recently launched to understand better the phenomena associated with the combustion hazard and to address the issues highlighted after the Fukushima Daiichi events such as explosion hazard in the venting system and the potential flammable mixture migration into spaces beyond the primary containment. The expected results will be used to improve the modeling tools and methodology for hydrogen risk assessment and severe accident management guidelines. The present paper aims to present the methodology adopted by Institut de Radioprotection et de $S{\hat{u}}ret{\acute{e}}$ $Nucl{\acute{e}}aire$ to assess hydrogen risk in nuclear power plants, in particular French nuclear power plants, the open issues, and the ongoing R&D programs related to hydrogen distribution, mitigation, and combustion.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3413-3420
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• 2021

Applying the available neutron flux for medical and industrial purposes is the most important application of research reactors. The neutron radiography system is used for non-destructive testing (NDT) of materials so that it is one of the main applications of nuclear research reactors. One of these research reactors is the 5 MW pool-type light water research reactor of Tehran (TRR). This work aims to investigate on materials and location of the beam tube (BT) of the TRR radiography system to improve the index parameters of BT. Our results showed that a through-type BT with 20 cm thick carbon neutron filter, 1.2 cm and 9.4 cm of the diameter of inlet (D₁) and output (D₂) BT, respectively gives thermal neutron flux almost 25.7, 5.6 and 1.1 times greater than the former design of the TRR (with D₁ = 1.8 cm and D₁ = 9.4 cm), previous design of the TRR with D₁ = 3 cm and D₁ = 9.4 cm, and another design with D₁ = 5 cm and D₁ = 9.4 cm, respectively. Therefore, the design proposed in this paper could be a better alternative to the current BT of the TRR.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2725-2732
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• 2020

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of ¹⁶O, ⁹Be, ²³⁵U, ²³⁸U, and S(𝛼,𝛽) of ¹H in H₂O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.9-18
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• 2024

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents' acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristics are analyzed through BALI experiments with similar condition. In this study, computational fluid dynamics (CFD) via Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) for BALI experiments were performed. Steady-state CFD analysis was performed on three turbulence models, and SST k-ω model was in good agreement with the experimental measurement temperature within the maximum error range of 1.9%. LES CFD analysis was performed based on the RANS analysis results and it was confirmed that the temperature and wall heat flux for depth was consistent within an error range of 1.0% with BALI experiment. The LES CFD analysis results were compared with those of the Lagrangian-based solver. LES matched the temperature distribution better than SOPHIA, but SOPHIA calculated the position of boundary between stratified layer and convective layer more accurately. On the other hand, Lagrangian-based solver predicted several small eddy behaviors of the convective layer and LES predicted large vortex behavior. The vibration characteristics near the cooling part of the BALI experimental device were confirmed through Fast Fourier Transform (FFT) investigation. It was found that the power spectral density for pressure at least 10 times higher near the side cooling than near the top cooling.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1689-1696
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• 2020

In this study, we developed a fiber-optic humidity sensor (FOHS) system for the monitoring and detection of coolant leakage in nuclear power plants. The FOHS system includes an FOHS, a spectrometer, a halogen white-light source, and a Y-coupler. The FOHS is composed of a humidity-sensing material, a metal tube, a multi-mode plastic optical fiber, and a subminiature version A (SMA) fiber-optic connector. The humidity-sensing material is synthesized from a mixture of polyvinylidene fluoride (PVDF) in dimethyl sulfoxide (DMSO) and hydroxyethyl cellulose (HEC) in distilled water. We measured the optical intensity of the light signals reflected from the FOHS placed inside the humidity chamber with relative humidity (RH) variation from 40 to 95%. We found that the optical intensity of the sensing probe increased linearly with the RH. The reversibility and reproducibility of the FOHS were also evaluated.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• The Transactions of the Korea Information Processing Society
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• v.4 no.6
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• pp.1550-1556
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• 1997

In this paper, we apply the evolutionary algorithm to the combinatorial optimization problem. Evolutionary algorithm useful for the optimization of the large space problem. This paper propose a method for the reuse of wastes of light water in atomic reactor system. These wastes contain several reusable elements, and they should be carefully selected and blended to satisfy requirements as an input material to the heavy water atomic reactor system. This problem belongs to an NP-hard like the 0/1 knapsack problem. Two evolutionary strategies are used as approximation algorithms in the highly constrained combinatorial optimization problem. One is the traditional strategy, using random operator with evaluation function, and the other is heuristic based search that uses the vector operator reducing between goal and current status. We also show the method which perform the feasible test and solution evaluation by using the vectored knowledge in problem domain. Finally, We compare the simulation results of using random operator and vector operator for such combinatorial optimization problems.

• Plant Journal
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• v.12 no.4
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• pp.37-43
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• 2016

Power system which provides electricity to the accident mitigation load for nuclear power plant should be verified to maintain the proper voltage level under the various loading and source conditions. For this purpose, it was needed to collect the voltage data of safety related buses during operation of the Reactor Coolant Pump(RCP) motor and Component Cooling Water Pump(CCWP) motor, respectively, under the certain loading condition of the plant. The data (such as, voltage, current, power factor) collected from actual measurement were used to modify the existing ETAP model and then the reanalysis was conducted to simulate the testing conditions. Through these actual measurement and analysis, it ensures that the existing electrical system analysis including assumptions and methods was conducted properly. Finally, the voltage of safety related buses was not dropped below the acceptable level, and the discrepancy between two results was within the limit.

• Journal of Environmental Health Sciences
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• v.37 no.3
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• pp.218-225
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• 2011

Objectives: In this paper, a dielectric barrier discharge (DBD) plasma reactor was investigated for degrading the dye Rhodamine B (RhB) in aqueous solutions. Methods: The DBD plasma reactor system in this study consisted of a plasma component [titanium discharge (inner), ground (outer) electrode and quartz dielectric tube], power source, and gas supply. The effects of various parameters such as first voltage (input power), gas flow rate, second voltage (output power), conductivity and pH were investigated. Results: Experimental results showed that a 99% aqueous solution of 20 mg/l Rhodamine B is decolorized following an eleven minute plasma treatment. When comparing the performance of electrolysis and plasma treatment, the RhB degradation of the plasma process was higher that of the electrolysis. The optimum first voltage and air flow rate were 160 V (voltage of trans is 15 kV) and 3 l/min, respectively. With increased second voltage (4 kV to 15 kV), RhB degradation was increased. The higher the pH and the lower conductivity, the more Rhodamine B degradation was observed. Conclusions: OH radical generation of dielectric plasma process was identified by degradation of N, N-dimethyl-4-nitrosoaniline (RNO, indicator of OH radical generation). It was observed that the effect of UV light, which was generated as streamer discharge, on Rhodamine B degradation was not high. Rhodamine B removal was influenced by real second voltage regardless of initial first and second voltage. The effects of pH and conductivity were not high on the Rhodamine B degradation.