Thermal-hydraulic Analysis Research Articles

Two-group drift-flux model for dispersed gas-liquid flows in rod bundles

Interfacial transfer terms in gas-liquid two-phase flows are formulated as the product of interfacial area concentration (IAC) and flux. The interfacial area transport equation (IATE) is essential for obtaining IAC in transient and developing two-phase flows. Two-group IATE was developed to account for the difference in the interfacial drag force between two groups of bubbles, where spherical and distorted bubbles are categorized as group one, while cap, slug, and churn turbulent bubbles are categorized as group two. The rigorous two-group approach requires two-group gas momentum equations, resulting in one additional momentum equation in the two-fluid model. The mixture gas momentum equation is considered to avoid the additional momentum equation. The two-group drift-flux model is necessary to calculate the two-group gas velocity from the mixture velocity. The present study develops an approximation methodology to acquire subchannel average and rod bundle average two-group two-phase flow parameters based on the validated power law assumption and a limited number of local data. The two-group model is developed for the distribution parameter and drift velocity for dispersed two-phase flows in rod bundles. The total performance of the newly developed two-group drift-flux model is evaluated by two-phase flow data in rod bundles. The results show that the developed two-group drift-flux model could well predict two-group gas velocities in rod bundles. This model is useful for subchannel thermal hydraulic analysis codes as a complement to the conventional one-group drift-flux model.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

Neutronics and thermal hydraulics analysis on a Lead-bismuth-cooled fast reactor

Breeding Lead-base Economical Safe System-Demonstration BLESS-D is a lead-bismuth-cooled fast reactor proposed by China State Power Investment Corporation Research Institute. It features a pool-type configuration of pressure vessel with a thermal/electric power of 300/100 MW. Neutronics and thermal hydraulics analysis are performed, mainly including a sensitivity study of the fuel components, fuel enrichment, and control-assembly arrangement on the lifetime, and the temperature filed for the fuel assembly with the largest thermal power. The results show that compared with PWR, the 239Pu production for BLESS-D is significantly higher, and the quantity of MAs is also reduced, while the productions of LLFPs are at the same scale. It has also been found that under steady state conditions, the largest temperatures of the fuel pellet, cladding and coolant are all below relevant criteria, showing a good performance in the thermal hydraulics for the design. Optimizing the fuel loading scheme for achieving a safer performance would be a part of the future’s work.

Preliminary thermal-hydraulic deterministic safety analysis of an in-vessel LOCA for the DTT facility

The DTT (Divertor Tokamak Test) facility is a new experimental tokamak under development by an Italian consortium in cooperation with several high-standard European laboratories. The ENEA division FSN-SICNUC is involved in the project for carrying out deterministic safety analysis of postulated accidents. In the first year of activity, it has been analyzed an In-Vessel LOCA (IVLOCA) scenario. The IVLOCA scenario selected is caused by a break in the divertor’s cassette cooling tubes with the consequent release of coolant inside the Vacuum Vessel (VV). The best estimate thermal-hydraulic system code TRACE, developed by USNRC, was selected to conduct the preliminary thermal-hydraulic analysis of this accident. DTT is currently under design, so not all the data are frozen. Therefore, based on some engineering assumptions and scaling considerations from similar facilities, the nodalization of DTT VV was developed using the three-dimensional TRACE component “VESSEL”. Then, a sensitivity analysis was carried out to simulate the break and the consequent water injection in the VV. This was done to compare the system behavior and test different nodalization approaches. Subsequently, the data of the DTT divertor cooling system were used to run additional simulations. The results allow to compare different nodalization approaches and to have a preliminary estimate of the pressure peak and temperature behavior in the VV for an IVLOCA. Finally, a first uncertainty analysis was carried out using the DAKOTA toolkit, coupled with TRACE code in SNAP. Two uncertain input parameters were selected: the rupture area of a cooling divertor tube and the temperature of the divertor coolant. The uncertainty analysis allows having a wider spectrum of system behavior and a preliminary insight on the dispersion of the VV pressure, selected as figure of merit. This paper aims to show the results of this preliminary analysis, characterizing the phenomena that occurred during the selected transient.

Estimating Allowable Operator Action Time for Mitigating Multiple‐Failure Accidents in Korean Advanced Power Reactor

In the case of multiple‐failure accidents in light water reactors, it is imperative to demonstrate that a significant fuel damage can be prevented via prearranged mitigation measures performed by operators. However, the time elapsed from the initiation of the event to the implementation of mitigation actions significantly influences the progress and consequences of the accident. Therefore, assuming a specific time for operator mitigation actions does not provide sufficient insights for safety assessments, thereby calling for further investigations into the effect of the time delay from the prescribed value. In this study, the allowable operator action time for multiple‐failure accidents of the Advanced Power Reactor 1400 was estimated by means of the sensitivity analysis using multi‐dimensional analysis for reactor safety‐KINS standard (MARS‐KS), a best‐estimate thermal–hydraulic analysis code used by the Korean nuclear regulatory authority. Three multiple‐failure accidents were investigated, including small‐break loss‐of‐coolant accident with the loss of safety injection (SI), intermediate‐break loss‐of‐coolant accident with the loss of SI, and total loss of feedwater accident. The upper limit of the mitigation action time, which prevents core damage, was estimated based on the requirement for peak cladding temperature. This was in turn used to derive the tolerable delay from the assumed action time in the Korean accident management program for each multiple‐failure accident.

Operating Reliability of In-grid HTS Power Cables Based on Electromagnetic–Thermal–Hydraulic Analysis

Operating Reliability of In-grid HTS Power Cables Based on Electromagnetic–Thermal–Hydraulic Analysis

Numerical analysis of reflood heat transfer and large-break LOCA including CRUD layer thermal effects

This study examined the effects of CRUD on reflood heat transfer behaviors of nuclear fuel rods during a loss-of-coolant-accident (LOCA) in a pressurized water reactor using a best-estimate thermal-hydraulic analysis code. Changes in thermal properties and boiling heat transfer characteristics of the CRUD layer were extensively reviewed, and a set of correction factors to reflect the changes was implemented into the code. A heat structure layer reflecting the effects of CRUDs on the properties was added to the outer surface of the fuel cladding. Numerical simulations were conducted to examine the effects of CRUDs on reflood cooling of overheated fuel rods for representative separate and integral effect tests, FLECHT-SEASET and LOFT. In LOFT analysis, the average cladding temperature was increased due to the low thermal conductivity of CRUD during steady-state operation; however, in both analyses, the peak cladding temperature decreased, and the quenching time was reduced. Obtained results revealed that when the porous CRUD layer is deposited on the fuel cladding, two opposite effects appear. Low thermal conductivity of the CRUD layer always increases fuel temperature during normal operation; however, its hydrophilic porous structures may contribute to accelerated reflood cooling of fuel rods during a LOCA.

Thermal-hydraulic analysis of lead-bismuth cooled reactors core based on single channel model

Thermal-hydraulic analysis of lead-bismuth cooled reactors core based on single channel model

Improved cryogenic stability by thermal insulation between forced-flow gas-cooled REBCO conductors in fusion magnets

A key design feature is presented to improve the cryogenic stability of forced-flow gas-cooled HTS conductors for fusion magnets. In order to realize the compact and economic fusion system, a variety of conductors are under development with stacked REBCO tapes so that the high-field magnets can operate at 20–30 K with the internal flow of helium gas. Thermal-hydraulic analysis is typically based on 1-D models under a prescribed thermal load along the conductors. In this study, the effect of thermal interaction between adjacent conductors in a winding pack is rigorously considered for the analysis. In a double pancake with single cooling loop, it is evidently verified that the conductor temperature may rise to a peak at inner layers, seriously affecting current sharing and thermal stability. The analytical results are useful in predicting the temperature and location of hot spot in terms of a dimensionless parameter, NTU. In theory, the gas-cooled magnet can be thermally stable by either infinitely increasing or infinitesimally decreasing the thermal interaction between conductors. In practice, however, the thermal insulation to minimize the contact between inbound and outbound flows in a double pancake is proposed as an effective way to improve the cryogenic stability.

Numerical simulation of thermal hydraulics of supercritical pressure water with 2×2 rod assembly wrapped differently with a wire

Through the use of CFD, a thermal-hydraulic analysis was carried out on a set of 4 rod assemblies. The common wire-wrapped spacers on the fuel assemblies are wound in the same direction for all rods. Because of this design, it is anticipated that there would be an adverse flow in all of the sub-channels. In the course of this analysis, this flow was demonstrated to be real. In order to achieve desirable flow characteristics for coolant mixing, a novel configuration for wire-wrapped spacers is proposed here. This pattern designates a group of 4 rods, the rods have the wrapping system of clockwise and anticlockwise pattern. It was determined that this pattern had an exceptional mixing effect as well as a uniform flow. This pattern has a maximum temperature area that is approximately 8.5 °C lower than that of the conventional wire-wrapped fuel bundle arrangement. When compared to the regular pattern, the pressure drop caused by the new arrangement was approximately 1.9% less severe. The redesigned layout of the rod assembly improves heat transfer while also reducing the amount of pressure loss that occurs.

Feasibility of Using Mixed-Oxide Fuel for a Pressurized Water Reactor Instead of Traditional UO2 Fuel Material

This study examines the feasibility of utilizing mixed-oxide fuel [(U0.9, rgPu0.1) O2] instead of traditional UO2 in nuclear reactors. The utilization of (U0.9, rgPu0.1) O2 is particularly significant as it represents an effective approach to nuclear fuel recycling by combining reactor-grade plutonium extracted from partially used nuclear fuel and depleted uranium obtained through the enrichment process. The fundamental neutronic characteristics, such as the radial power distribution, were investigated using the MCNPX 2.7 algorithm to identify the specific channel for subsequent thermal-hydraulic (TH) analysis. The TH analysis was conducted using COMSOL-Multiphysics, allowing for the estimation of the fuel rod’s axial and radial temperature profiles, as well as the determination of the departure from the nucleate boiling ratio. Furthermore, the coupling between heat transfer and solid structure (SS) was achieved using the Multiphysics tool in COMSOL-Multiphysics. This coupling facilitated the simulation of key SS parameters, including von Mises stress, volumetric strain, and displacement, while considering the influence of heat transfer. The results demonstrate significant improvements and enhanced safety margins when utilizing (U0.9, rgPu0.1) O2.

CFD analysis of PHWR exposed core under postulated severe accident condition

A two-dimensional steady-state thermal-hydraulic analyses of the exposed core with varying levels of exposure is carried out for heat transfer assessment. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in Pressurised Heavy Water Reactors (PHWRs). Such exposure of fuel channels leads to their heat-up which is governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer. In order to understand such a complex behaviour as well as to plan the severe accident management guidelines, numerical analyses of the exposed core are required. In this paper, one-of-a-kind thermal-hydraulic analyses of the exposed reactor-scale core geometry are presented for row-wise exposures of fuel channels. These numerical analyses are terminated at the onset of fuel channel disassembly i.e., loss of geometrical intactness of the exposed core. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is brought out in this study. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, for the advanced stages of fuel channel exposures, the heat transfer is observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation arising from Zircaloy and steam interaction. Moreover, it is interesting to note that the disassembly of fuel channels is restricted to central regions of the exposed core.

Thermal-hydraulic analysis of LBE flow in rod bundles with wire spacers based on heat transfer framework of OpenFOAM

The wire spacer allows for a compact layout in the core of lead-cooled fast reactor (LFR). However, it also introduces complexity to the flow and heat transfer of the coolant, lead-bismuth eutectic (LBE). This paper aims to analyze the effect of wire spacer in rod bundle and the applicability of heat transfer framework to LBE flow. Three Reynolds-Averaged Navier–Stokes (RANS) turbulence models with two representative turbulent Prandtl number (Prt) 1.5 and 2.0 are employed within the framework of OpenFOAM. The sensitivity of the models is examined based on turbulent heat transfer in tube flow. The simulations considered with an extensive mesh sensitivity study were analyzed. The buoyantSimpleFoam solver is compared with direct numerical simulation (DNS) data. Thermal-hydraulic analysis are conducted, considering temperature, velocity distribution, and local Nusselt (Nu) number, and pressure drop based on 7-rod and 19-rod bundle with wire spacers. The influence of wire initial starting angles from 0°, 10°, 20°, 30°, 40°, 50° on heat transfer is carried out. The study reveals that an optimum angle range from 20° to 40° exists in the sub-channels. This work validates the basic heat transfer solver for LBE using OpenFOAM.

Numerical study on a 7-pin bundle fuel assembly with different wire arrangements

The sodium cooled fast reactor is a Gen-IV reactor with great development prospect. The thermal-hydraulic analysis of the fuel assembly is crucial to the safety and efficiency of the reactor operation. In this paper, four assemblies with different wire arrangements were established to study the influence of wire arrangement on the thermal-hydraulic characteristics of the fuel assembly using Computational Fluid Dynamics (CFD). The solver is steady state and flow convergence is taken into account when the residuals for all flow variables are less than 10−5. It is observed that among these four arrangements, the reverse arrangement exhibits the largest average transverse flow and global Nusselt number. Additionally, the temperature difference between different sub-channels is the smallest, resulting in the most uniform temperature field. Furthermore, the average temperature of the fuel cladding is notably decreased in comparison to other arrangements. To evaluate the pressure loss and heat transfer performance of the assembly, the performance evaluation criterion PEC is introduced. The findings reveal that the PEC of the reverse arrangement is superior; in other words, it simultaneously reduces pressure drop and enhances heat transfer.

Spanish research related to SMRs projects

Small modular reactors (SMRs) are advanced nuclear reactors with a power capacity of up to 300 MW(e) per unit. SMRs encompass a variety of reactor technologies including light water reactors, high temperature gas reactors, molten salt reactors, liquid metal cooled fast reactors, and heat pipe technology-based reactors. The research and design of these diverse SMR types require a broad set of technological capabilities related to nuclear engineering and safety. Within this context, this article attempts to assess the current state of research and technological progress achieved by the Spanish research groups and companies. The results reveal a significant level of maturity among these groups and companies in various domains such as neutronic analysis, thermal hydraulic analysis, the improvement of models related to severe accident scenarios and an active involvement in the design of novel SMR.

Thermal-Hydraulic comparative analysis of G-III + BWR and PWR passive safety features under Loss of coolant accident

As one of symbolic design features of Generation-III and III+ (G-III and G-III + ) nuclear reactors, the passive safety system has been implemented on majority of G-III and III + PWRs and BWRs. The passive safety systems in those reactors are designed differently yet adhere to identical fundamental principles. It would be interesting to examine the similarities and differences between the passive safety systems of G-III + PWR and BWR. System code simulations for Small Break and Large Break Loss of Coolant Accident for the Economic Simplified Boiling Water Reactor (ESBWR) and China’s advanced PWR had been conducted, respectively. The results showed that the passive safety systems of ESBWR and China’s Advanced PWR were comparable in design, and when dealing with the similar type of SBLOCA and LBLOCA, the passive safety components with the same functionality could all contribute to the core's cooling with stability and reliability, and the injection flow rate values were nearly equivalent. Separating the injection lines of ESBWR's safety systems guarantees that the action sequence could well remain stable. This design could prevent the injection flow rate and action sequence from being impacted by the interaction of safety injection systems. This work could serve as a reference for the design of subsequent new reactors, thereby contributing to the design of reactors with enhanced economics and safety.

Load study and structural analysis for CFETR divertor OVT

China Fusion Engineering Test Reactor (CFETR) is an important device for studying fusion reactor engineering in China. The Outer Vertical Target (OVT) are high heat load components of the divertors and work in a strong electromagnetic field, so they need sufficient heat removal capability and structural strength. In addition, due to activation of neutrons on materials, the OVT also needs to meet the requirements of Remote Handling (RH) maintenance. CFETR puts forward the concept of hybrid divertor-blanket, which cancels the cassette and adopts a separate structure. The lower part of the divertor is occupied by the breeding blanket, and the divertor target is fixed on the breeding blanket, to improve the RH maintenance efficiency and Tritium Breeding Ratio (TBR). In previous work, the structural design and thermal-hydraulic analysis of the OVT were completed. In this paper, load study and structural analysis was carried out to verify the structural strength of the OVT under experimental test conditions (including liquid pressure, nitrogen baking), normal operation conditions (including the impact of thermal stress under thermal load, gravity, working pressure, etc.) and extreme event conditions (Vertical Displacement Event (VDE) or Major Disruption (MD) event, including the impact of thermal stress under thermal load, gravity, working pressure, electromagnetic load, etc.). The results show that the OVT meets the requirements of heat removal and strength, and effectively promote the processing and manufacturing of the prototype of the OVT.

Preliminary conceptual design with neutronics and thermal-hydraulics assessments of FuSTAR

The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction.

Phenomena Identification and Ranking Table for Thermal-Hydraulic Safety of China Fusion Engineering Test Reactor

The China Fusion Engineering Test Reactor (CFETR) is presently in the engineering design phase, and it is crucial to establish a thermal-hydraulic safety analysis method. First of all, the identification of phenomena in selected accident scenarios affecting reactor safety performance metrics should be focused on with the phenomenon identification and ranking table (PIRT), in which phenomena related to the thermal-hydraulic safety of the vacuum vessel and related systems are screened, including both design-basis accidents and beyond-design-basis accidents. The development of this PIRT exercise for CFETR thermal-hydraulic safety is addressed in this study; specifically, the importance ranking and stage of knowledge (SoK) of phenomena in the selected accidents are evaluated by the PIRT panel. The results of the PIRT analysis revealed that there exist certain safety-critical phenomena for which SoK is relatively low despite their perceived significance in safety performance metrics, which include the phase change of coolant, the migration of the multicomponent gas mixture, the migration of radionuclides within confinement, and the dispersion of gaseous radionuclides in the atmosphere. Ongoing research works and follow-up plans to improve the SoK of phenomena are presented.

A subchannel analysis code SACOS-Na for sodium-cooled fast reactor

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm.For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes.