• Energy Research

Abstract It is necessary to accurately predict the minimum point of pressure drop to ensure the safety of nuclear reactors. However, the non-uniform heat flux distribution along the transverse direction is encountered when the plate-type nuclear fuels are used. This study shows the effect of a transversely non-uniform heat flux on the minimum point of the pressure drop. The pressure drop-flow rate curve under the non-uniform heat flux was obtained by the experiment, and the trend of curve was different with the one of uniform heat flux case. Under the non-uniform heat flux, even when the inlet mass flow rate decreased, the value of the pressure drop was constant for a while with the development of a two-phase flow. With further reduction of inlet mass flow rate, the pressure drop started to decrease until the minimum point of the pressure drop was reached. Moreover, the inlet mass flow rate at the minimum point of pressure drop is much lower than that in the uniform heat flux case. For a detail analysis, the numerical approach is proposed along with the application of multi-channel concept. A single narrow rectangular channel is divided along the transverse direction, and the heat flux is given non-uniformly to the divided channels. Although the pressure drop is separately calculated for each divided channel, the mass is transferred between the channels. In the calculation, the mass flow rate is non-uniformly distributed in the transverse direction. If the mass flow rate is uniformly distributed, the non-uniform heat flux causes an unbalanced pressure drop because of the non-uniform distribution of void fraction. As a result, at the edges where the void fraction is high, the mass flow rate is transferred to the middle of channel to balance the pressure drop in transverse direction. When the void fraction in the middle becomes significantly large, the minimum point of the pressure drop can be obtained.

Abstract A code-to-code comparison study was performed to investigate the thermal hydraulic characteristics of the JRR-3 reactor. The COOLOD-N2 and TMAP codes were used to analyze the JRR-3 reactor under downward forced convective and natural convective flows. The geometry and operational conditions of JRR-3 used in the present study were provided by the NEA (Nuclear Energy Agency) data bank. The total power peaking factors such as the radial and axial power peaking and the engineering hot channel factors were implemented to simulate hot and average channels. Thermal hydraulic characteristics, i.e., coolant and cladding wall surface temperatures, and the flow velocity and thermal margins, i.e., minimum ONB (Onset of Nucleate Boiling) temperature margin and minimum DNB (Departure from Nucleate Boiling) ratio were estimated at a nominal power of 20 MW for forced convection cooling and at 0.2 MW for natural convection cooling. As a result, a comparison study between the COOLOD-N2 and TMAP codes showed a comparability of the thermal hydraulic analysis results except for the ONB temperature margin and the coolant velocity in the hot channel during natural convection cooling. The differences resulted from the present analyses were discussed.

Abstract This study investigates the effect of transverse power distribution on the ONB (Onset of Nucleate Boiling) incipient. For this purpose, a subcooled boiling model with uniform and non-uniform heat flux distribution is simulated in a narrow vertical rectangular channel heated from both sides by applying a wide range of thermal power (8–16 kW). The simulations are performed using the CFX and TMAP codes. The CFX code incorporates both a two-fluid model and RPI wall boiling model to investigate coolant and wall temperature distributions along the heated channel. The TMAP code implements two different sets of heat transfer correlations to evaluate the wall temperature. The results obtained from the TMAP analyses show that the wall temperatures predicted by the Jo et al. heat transfer correlation are higher than the ones predicted by the Dittus and Boelter heat transfer correlation. The wall temperatures predicted by the CFX analyses lie between the predicted wall temperatures obtained by the TMAP analyses. Based on the superheated temperature on the heated surface, the ONB incipient is determined. The axial locations of the ONB incipient are predicted differently by the CFX and TMAP analyses. For uniform heating, the ONB incipient predicted by the CFX analysis occurs between the predictions made by the TMAP analyses. For non-uniform heating, the ONB incipient by the CFX analysis occurs at a higher power than the power required by the TMAP analyses.

Heat transfer characteristics in a narrow rectangular channel are experimentally investigated for upward and downward flows. The experimental data obtained are compared with existing data and predictions by many correlations. Based on the observations, there are differences from others: (1) there are no different heat transfer characteristics between upward and downward flows, (2) most of the existing correlations under-estimate heat transfer characteristics, and (3) existing correlations do not predict the high heat transfer in the entrance region for a wide range of Re. In addition, there are a few heat transfer correlations applicable to narrow rectangular channels. Therefore, a new set of correlations is proposed with and without consideration of the entrance region. Without consideration of the entrance region, heat transfer characteristics are expressed as a function of Re and Pr for turbulent flows, and as a function of Gz for laminar flows. The correlation proposed for turbulent and laminar flows has errors of ±18.25 and ±13.62%, respectively. With consideration of the entrance region, the heat transfer characteristics are expressed as a function of Re, Pr, and z* for both laminar and turbulent flows. The correlation for turbulent and laminar flows has errors of ±19.5 and ±22.0%, respectively.

Abstract Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature.

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

We validated the performance of RELAP MOD3.3 code regarding the hybrid SIT with available experimental data. The concept of the hybrid SIT is to connect the pressurizer to SIT to utilize the water inside SIT in the case of SBO or SB-LOCA combined with TLOFW. We investigated how well RELAP5 code predicts the physical phenomena in terms of the equilibrium time, stratification, condensation against Separate Effect Test (SET) data. We also conducted the validation of RELAP5 code against Integrated Effect Test (IET) experimental data produced by the ATLAS facility. We followed conventional approach for code validation of IET data, which are pre-test and post-test calculation. RELAP5 code shows substantial difference with changing number of nodes. The increase of the number of nodes tends to reduce the condensation rate at the interface between liquid and vapor inside the hybrid SIT. The environmental heat loss also contributes to the large discrepancy between the simulation results of RELAP5 and the experimental data.

Abstract Onset of Nucleate Boiling (ONB) is an important limit in the design of nuclear reactors and most flow boiling systems. Preventing the ONB occurrence protects systems from unfavorable thermal hydraulic events, such as Onset of Flow Instability (OFI) and Critical Heat Flux (CHF). In this study, a simultaneous measurement and visualization experiment on the ONB is carried out for a narrow rectangular channel heated from one side. The rectangular channel has a thickness of 2.35 mm, width of 54 mm, and length of 560 mm. The experiment is conducted for upward flow direction under nearly atmospheric pressure. The inlet conditions are chosen to cover a wide range of operational conditions: inlet temperature (35–65 °C) and mass flow rate (0.015–0.130 kg/s). Based on the inlet flow conditions, a uniform heat flux (50–800 kW/m2) is applied in a stepwise manner to the heated surface. The slope of the wall temperature versus the heat flux curve decreases at the ONB point. Ten thermocouples (TCs) are installed into the heated block to measure the wall temperature distribution. On the other hand, a high-speed camera is used to visualize the ONB point and compare it with the wall temperature deviation point. Based on the experimental data, the influence of mass flow rate and inlet temperature on the bubble behavior along the test section is observed. The results show a new trend for the influence of inlet temperature on the superheated temperature of the wall at the ONB point. A similar trend is observed using CFX analysis for the test section. The present results are compared with other experimental studies conducted by different research institutes, with different ONB heat flux correlations such as in the studies of Jens and Lottes (1951), Bergles and Rohsenow (1964), and Thom et al. (1965). The correlations underestimate the experimental results. Therefore, a new correlation is developed to predict the ONB heat flux, which has good agreement with the experimental data within an error of ±16.5%.