• Progress in Superconductivity and Cryogenics
• /
• v.21 no.2
• /
• pp.22-25
• /
• 2019

The superconducting magnetic separation system has been developing to separate the iron oxide scale from the feed water of the thermal power plant. The accumulation in the boiler lowers the heat exchange rate or in the worst case damages it. For this reason, in order to prevent scale generation, controlling pH and redox potential is employed. However, these methods are not sufficient and then the chemical cleaning is performed regularly. A superconducting magnetic separation system is investigated for removing iron oxide scale in a feed water system. Water supply conditions of the thermal power plant are as follows, flow rate 400 t / h, flow speed 0.2 m / s, pressure 2 MPa, temperature $160-200^{\circ}C$, amount of scale generation 50 - 120 t / 2 years. The main iron oxide scale is magnetite (ferromagnetic substance) and its particle size is several tens ${\mu}m$. As the first step we are considering to introduce the system to the chemical cleaning process of the thermal power plant instead of the thermal power plant itself. The current status of development will be reported.

• Corrosion Science and Technology
• /
• v.18 no.4
• /
• pp.138-147
• /
• 2019

Pipes and components of the secondary system in the pressurized water reactor (PWR) are mainly comprised of manufactured carbon steel. Thus, the generated carbon steel corrosion products are transported into the steam generator and deposited, thereby deteriorating the integrity of the steam generator. Environmental condition in the secondary system of the PWRs differs across different locations. So, the corrosion rate and types of corrosion products depend on specific locations in the secondary system. In this study, the quantity and chemical compositions of corrosion products generated in various locations that vary in different temperatures and chemistry conditions were investigated. As a result of evaluating the PWR "Unit A" that is in current operation, the amount of corrosion products generated in the section of high temperature feedwater system was identified as the largest source in the secondary system. Major components of corrosion products were iron oxides such as magnetite, hematite, and lepidocrocite.

• Nuclear Engineering and Technology
• /
• v.54 no.12
• /
• pp.4601-4619
• /
• 2022

The classical Probabilistic Safety Analysis (PSA) does not include any time dependence explicitly. However, the success criteria (SC) could evolve during the cycle for some initiating events. In that sense, there is a type of sequence in which this time-dependency is quite important, the family of Anticipated Transient without Scram (ATWS) sequences in Pressurized Water Reactors. Therefore, a new risk-informed approach is proposed in this paper, which makes it possible to obtain the time-dependent SC evolution of the safety functions affected by the Moderator Temperature Coefficient (MTC) value. Then, the evolution of the ATWS conditional core damage probability (CCDP) could be obtained using a PSA model. To quantify the CCDP, the average values of the time-dependent failure probabilities must be computed. Finally, the comparison between the CCDP obtained through the application of the classical PSA approach and the new one makes it possible to quantify the impact of time-dependence on the SC of the headers that this new risk-informed ATWS approach can provide.

• Nuclear Engineering and Technology
• /
• v.18 no.2
• /
• pp.97-106
• /
• 1986

System thermal-hydraulic parameters and simulated, using the best-estimate system code(RELAPS/MODl/NSC), based upon the sequence of events for the KNU1 (Korea Nuclear Unit 1) loss of offsite power transient at 77.5% power which occurred on June 9,1981. The results are compared with the actual plant transient data and show good agreements. After the flow coastdown following the trips of both reactor coolant pumps, the establishment of natural circulation by the temperature difference between the hot and the cold legs is confirmed. The calculated reactor coolant flowrate closely approximates the plant data indicating the validity of relevant thermal-hydraulic models in the RELAP5/MOD1/NSC. Results also show that the sufficient heat removal capability is secured by the appropriate supply of the auxiliary feedwater without the operation of S/G PORVs. In addition, a scenario accident at full power, based upon the same sequence of events described above, is also analysed and the results confirmed that the safety of KNU1 is secured by the appropriate operation of the S/G PORVs coupled with the supply of auxiliary feedwater which ensures sufficient heat removal capability. The characteristics of the non-safety related components such as the turbine stop valve closing time, S/G PORV settings etc. are recognized to be important in the transient analyses on a bestestimate basis.

• Environmental Engineering Research
• /
• v.15 no.2
• /
• pp.79-84
• /
• 2010

Three biological aerated filters (BAFs) composed of a PVC pipe with a diameter of 75 mm were constructed and operated at a waste-water temperature at $13^{\circ}C$. The media used for each BAF were: 5-mm gravel; 5-mm lava rock; 12.5-mm diameter by 15-mm long plastic rings, all with a media depth of 1.7 m. The feedwater, which simulated the effluent of aerated lagoons, had influent soluble chemical oxygen demand (sCOD) and ammonia concentrations of approximately 50 and 25 mg/L, respectively. For a hydraulic retention time (HRT) of two hours without recirculation, ammonia percent removals were 98.5, 98.9, and 97.8%, for the gravel, lava rock, and plastic rings, respectively. By increasing the effluent recirculation from 100 to 200% for an HRT of one hour, respective ammonia removals improved from 90.1 to 96, 76.5 to 90, and 65.3 to 79.5% for gravel, lava rock, and plastic rings. Based on the ammonia and sCOD loadings for different HRTs, the estimated maximum ammonia loading was approximately 0.6 kg $NH_3-N/m^3$-day for the three BAFs of different media types. The zero-order biotransformation rates for the BAF with gravel were found to be higher than the lava rock and plastic ring media. The results ultimately showed that BAF can be used as an add-on system to aerated lagoons or as a secondary treatment unit to meet ammonia discharge limits.

• The KSFM Journal of Fluid Machinery
• /
• v.17 no.6
• /
• pp.109-114
• /
• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Nuclear Engineering and Technology
• /
• v.19 no.1
• /
• pp.10-21
• /
• 1987

Uncertainty analysis of the FLB accident is performed for KNU-1 using the response surface methodology and Monte Carlo simulation. The FLB analyses using the RELAP4/Mod6 were performed a number of times to generate the data base for the uncertainty analysis, along with the EM calculation for comparison purpose. Two kinds of input sets are utilized for response surface method to investigate and compare the effects of the uncertainty of input variables on the RCS peak pressure following a FLB. The first set is composed of six major plant operational parameters and the second set is composed of five major modelling parameters. It is found through the analysis of results that the uncertainties of modelling parameters have more influence on the RCS peak pressure than the uncertainties of plant operational parameters and that the extra margin of 9% of peak pressure is gained. And one of the assumptions of EM calculation, which is usually accepted as conservative is found to be erroneous, that is, the initial core inlet temperature is found to act negatively on the RCS pressure following a FLB.

• Membrane Journal
• /
• v.29 no.6
• /
• pp.314-322
• /
• 2019

A worsened drought in Chungnam province of Korea due to climate change and increasing water demand at Daesan industrial complex have motivated the 100,000 ㎥/d seawater desalination project. In this study, therefore, the comprehensive analysis of parameters affecting the reverse osmosis (RO) performance was conducted. Under the various conditions of feedwater salinity and temperature in Daesan, energy consumption was calculated as 2.39 ± 0.13 kWh/㎥. The decrease in membrane flux and recovery rate positively impacted annual operation cost. The average total dissolved solids (TDS) of the permeate and energy consumption with highly permeable membrane according to the membrane manufacturer were 3.84 mg/L and 2.22 ± 0.13 kWh/㎥, respectively. In addition, energy saving up to 0.29 kWh/㎥ or cost reduction of membrane module up to 15.6% is expected by changing the RO configuration from full two pass to partial or split partial two pass.

• Nuclear Engineering and Technology
• /
• v.50 no.3
• /
• pp.356-367
• /
• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.