• Nuclear Engineering and Technology
• /
• v.30 no.4
• /
• pp.318-327
• /
• 1998

The deuterium uptake behavior of Zr-2.5Nb pressure tubes in Wolsong Unit 1 was analyzed in terms of longitudinal location, operation time, and coolant temperature. The results were compared with those obtained from Canadian CANDU reactors. The amount of deuterium uptake was higher at the outlet part than at the inlet part and was also higher when subjected to a longer operation time and a higher coolant temperature. The hydrogen uptake of Zr-2.5Nb in a hydrogen gas atmosphere was dependent on the microstructure of the alloy. The aged Zr-2.5Nb consisting of $\alpha$-Zr and $\beta$-Nb phases showed higher hydrogen uptake than that consisting of $\alpha$-Zr and $\beta$-Zr phases. The hydrogen in the alloy decreased the rate of oxidation. This could be explained in terms of the cathodic controlled reaction of Zr-2.5Nb oxidation.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.29 no.12
• /
• pp.654-662
• /
• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• Journal of Power System Engineering
• /
• v.9 no.1
• /
• pp.14-22
• /
• 2005

This work deals with a controlled auto-ignition (CAI) single cylinder gasoline engine, focusing on the extension of operating conditions. The fuel is injected indirectly into electrically heated inlet air flow. In order to keep a homogeneous air-fuel mixing, the fuel injector is water-cooled by a specially designed coolant passage. Investigated are the engine performance and emission characteristics under the wide range of operating conditions such as 40 in the air-fuel ratio, 1000 to 1800 rpm in the engine speed, $150\;to\;180^{\circ}C$ in the inlet-air temperature, and $80^{\circ}$ BTDC to $20^{\circ}$ ATDC in the injection timing. A controlled auto-ignition gasoline engine can be achieved that the ultra lean-burn with self-ignition of gasoline fuel by heating inlet air. It can be achieved that the emission concentrations of carbon monoxide, hydrocarbons and nitrogen oxide had been significantly reduced by CAI combustion compared with conventional spark ignition engine.

• Journal of the Korean Society of Manufacturing Process Engineers
• /
• v.21 no.6
• /
• pp.44-49
• /
• 2022

The improvement in the battery performance and life using a battery thermal management system directly affects the improvement in the performance, life, and energy efficiency of electric vehicles. Therefore, this study numerically analyzed the heat exchange processes between the coolant inside the cooling plate channel and the heat generated by the battery. The cooling performance was analyzed based on the average temperature, temperature uniformity, and the maximum and minimum temperature differences of the battery. A performance difference existed depending on the coolant inlet temperature but showed the same tendency of cooling performance according to the shape of each plate's channel. Type 1 showed the best results in terms of battery temperature uniformity, which is the most important measure of battery performance; Type 2 showed the best results in terms of the average temperature of the battery; and Type 3 showed the best results in terms of the maximum and minimum temperature differences of the battery compared with that of the other cooling plates.

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2699-2708
• /
• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.11 no.1
• /
• pp.18-30
• /
• 1999

A heat transfer analysis for the two-dimensional (2-D) steady state using finite difference method (FDM) is performed to predict the thermal behavior of the primary first-wall (FW) system of fusion reactor under various geometric and thermo-hydraulic conditions, such as the beryllium (Be) armor thickness, pitch of cooling tube, and coolant velocity. The FW consists of authentic steel (type 316 stainless steel solution annealed) for cooling tubes, Cu for cooling tubes embedding material, and Be for a protective armor, based on the International Thermonuclear Experiment Reactor (ITER) report. The present 2-D analysis, the control volume discretized with hybrid grid (rectangular grid and polar grid) and Gauss-Seidel iteration method are adapted to solve the governing equations. In the present study, geometric and thermo-hydraulic parameters are optimized with consideration of several limitations. Consequently, it is suggested that the adequate pitch of cooling tube is 22-32mm, the beryllium armor thickness is 10-12mm, and that the coolant velocity is 4.5m/s-6m/s for $100^{\circ}C$ of inlet coolant temperature. The cooling tube should locate near beryllium armor. But, it would be better for locating the center of Cu wall, considering problems of material and manufacturing. Also, 2-D analysis neglecting the axial temperature distribution of cooling tube is appropriate, regarding the discretization error in axial direction.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.26 no.5
• /
• pp.224-230
• /
• 2014

The purpose of this paper is to investigate the performance of supplemental cooling and heating system equipped with the 1 kW thermoelectric module. The system consist of 96 thermoelectric modules, heat sink with louver fin and water cooling jacket which is attached on the hot side of the thermoelectric module. The cooling and heating performance test of the thermoelectric system is conducted with various conditions, such as intake voltage, air inlet temperature, air flow volume, water inlet temperature and water flow rate at calorimeter chamber in consideration of environmental conditions in realistic vehicle drive. The experimental results of a thermoelectric system shows that the cooling capacity and COP is 1.03 kW, and 1.0, and heating capacity and COP is 1.53 kW, and 1.5 respectively.

• Journal of computational fluids engineering
• /
• v.20 no.1
• /
• pp.84-91
• /
• 2015

SMART adopts, very unique facility, an FMHA to enhance the thermal and flow mixing capability in abnormal conditions of some steam generators or reactor coolant pumps. The FMHA is important for enhancing thermal mixing of the core inlet flow during a transient and even during accidents, and thus it is essential that the thermal mixing characteristics of flow of the FMHA be understood. Investigations for the mixing characteristics of the FMHA had been performed by using experimental and CFD methods in KAERI. In this study, the temperature distribution at the core inlet region is investigated for several abnormal conditions of steam generators using the commercial code, FLUENT 12. Simulations are carried out with two kinds of FMHA shapes, different mesh resolutions, turbulence models, and steam generator conditions. The CFD results show that the temperature deviation at the core inlet reduces greatly for all turbulence models and steam generator conditions tested here, and the effect of mesh refinement on the temperature distribution at the core inlet is negligible. Even though the uniformity of FMHA outlet hole flow increases the thermal mixing, the temperature deviation at the core inlet is within an acceptable range. We numerically confirmed that the FMHA applied in SMART has an excellent mixing capability and all simulation cases tested here satisfies the design requirement for FMHA thermal mixing capability.

• 한국전산유체공학회:학술대회논문집
• /
• 2008.03b
• /
• pp.292-295
• /
• 2008

In a SCWR (SuperCritical pressure Water cooled Reactor), the coolant temperature initially at below the pseudo-critical temperature at the bottom of a reactor core increases as the coolant flows upward through the sub-channels of the fuel assemblies, and it finally becomes higher than the pseudo-critical temperature when it leaves the reactor core. At certain conditions, heat transfer deterioration occurs near the pseudo-critical temperature and it may cause a drastic rise of the fuel surface temperature resulting a fuel failure. Therefore, an accurate estimation of the heat transfer coefficient is very important for the thermal-hydraulic design of a reactor core. An experiment on heat transfer to the vertically upward flowing $CO_2$ at a supercritical pressure in a circular tube were performed at KAERI. The internal diameter of the test section is 6.32 mm, which corresponds to the hydraulic diameter of a sub-channel in the conceptional design proposed by KAERI. The test range of the mass flux is 285 to 1200 kg/m$^2$s and the maximum heat flux is 170 kW/m$^2$. The inlet pressure is maintained at 8.12 MPa, which is 1.1 times the critical pressure. A new correlation, which covers both the normal and deterioration heat transfer regimes was proposed and compared with the estimations by exiting correlations.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3182-3195
• /
• 2021

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.