Hydrogen is adopted as coolant for regenerative cooling nozzle and reactor core in nuclear thermal propulsion (NTP), which is a promising technology for human space exploration in the near future due to its large thrust and high specific impulse. During the cooling process, the hydrogen alters its state from subcritical to supercritical, accompanying with great variations of fluid properties and heat transfer characteristics. This paper is intended to study heat transfer processes of supercritical pressure hydrogen under extremely high heat flux by using numerical approach. To begin with, the models explaining the variation of density, specific heat capacity, viscosity, and thermal conductivity are introduced. Later on, the convective heat transfer to supercritical pressure hydrogen in a straight tube is investigated numerically by employing a computational model, which is simplified from experiments performed by Hendricks et al. During the simulation, the standard k– ε model combining the enhanced wall treatment is used to formulate the turbulent viscosity, and the results validates the approach through successful prediction of wall temperature profile and bulk temperature variation. Besides, the heat transfer deterioration which may occur in the heat transport of supercritical fluids is also observed. According to the results, it is deduced that the flow acceleration to a flat velocity profile in the near wall region due to properties variation of hydrogen contributes to the suppression of turbulence and the heat transfer deterioration, while the “M-shaped” velocity profile is more often correlated to the starting of a recovery phase of turbulence production and heat transfer.

Subchannel thermal-hydraulics program named CORe thermal-hydraulics (CORTH) and assembly lattice calculation program named KYLIN2 have been developed in Nuclear Power Institute of China (NPIC). For the sake of promoting the efficiencies of these programs and achieving the better description on fined parameters of reactor, programs' linear systems and details are interpreted and parallelized. Test results show that the calculation efficiencies of linear systems occupy a large proportion of according serial computation. Based on the analysis, both programs' efficiencies are improved greatly through the proposed distributed-memory parallel strategy.

One-region (1-R) sensitivity computations with the annular-flow model were carried out for countercurrent flow limitation (CCFL) at a sharp-edged lower end in vertical pipes to generalize the prediction method for CCFL there (i.e., predicting effects of diameters and fluid properties on CCFL characteristics). In our previous study, we selected a correlation of interfacial friction coefficients, f i , with a function of average void fraction which gave a good prediction of the trend for air–water CCFL data, and we modified it to get good agreement with steam–water CCFL data under atmospheric pressure conditions, but it failed to predict CCFL reasonably at high pressure conditions. We recently found a Russian report on CCFL data at high pressure conditions, by which we improved the f i correlation using the dimensionless diameter and the viscosity ratio or density ratio of gas and liquid phases to get good agreement with CCFL data at high pressures. The improved f i correlation with the viscosity ratio and the improved f i correlation with the density ratio gave similar computed results, but the number of adjustment functions was one for the density ratio and two for the viscosity ratio (i.e., minimum value of two functions).

The effective thermal diffusivity and conductivity of pebble bed in the high temperature gas-cooled reactor (HTGR) are two vital parameters to determine the operating temperature and power in varisized reactors with the restriction of inherent safety. A high-temperature heat transfer test facility and its inverse method for processing experimental data are presented in this work. The effective thermal diffusivity as well as conductivity of pebble bed will be measured at temperature up to 1600 °C in the under-construction facility with the full-scale in radius. The inverse method gives a global optimal relationship between thermal diffusivity and temperature through those thermocouple values in the pebble bed facility, and the conductivity is obtained by conversion from diffusivity. Furthermore, the robustness and uncertainty analyses are also set forth here to illustrate the validity of the algorithm and the corresponding experiment. A brief experimental result of preliminary low-temperature test is also presented in this work.

The present experimental investigation in a scaled facility of an Indian pressurized heavy water reactors (PHWRs) is focused on the heat transfer behavior from the calandria vessel (CV) to the calandria vault during a prolonged severe accident condition in the presence of decay heat. The transient heat transfer simulates the conditions from single phase to boiling in the calandria vault water, partial uncovery of the CV due to boil off of water in the vault, and refill of calandria vault. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1100 °C. Decay heat in the melt pool was simulated by using high watt cartridge type heaters. The temperature distributions inside the molten pool across the CV wall thickness and vault water were measured for prolonged period which can be divided into various phases, viz., single phase natural convection heat transfer in calandria vault, boiling heat transfer in calandria vault, partial uncovery of CV, and refilling calandria vault. Experimental results showed that once the crust formed, the inner vessel temperature remained very low and vessel integrity maintained. Even boiling of calandria vault water and uncovery of CV had negligible effect on melt, CV, and vault water temperature. The heat transfer coefficients on outer vessel surface were obtained and compared with various conditions.

This study investigates the reactor core physical properties of the AP1000 ® , which applies the MCNP4a program to model the AP1000 reactor core with the parameters and data from the design control document (DCD, Rev. 19) of the AP1000 Nuclear Power Plant, which has been submitted to the nuclear regulatory commission (NRC). The model is applied to calculate and verify the physical parameters of AP1000 core design. The results match well with the design values in the DCD of the AP1000 nuclear power plant. The model will be modified according to the actual reactor core arrangement, such as AP1000 reactors at China's Sanmen and Haiyang sites, and then compared with the commissioning test results in the future.

Nuclear fuel rods operate under complex radioactive, thermal, and mechanical conditions. Nowadays, fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three-dimensional (3D) module within the framework of a general-purpose finite element solver, i.e., abaqus , for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermomechanical behavior of fuel rods. The swelling of fission products causes a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress in the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.

During a severe accident (SA) in a nuclear power plant (NPP), there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system (FCVS) can be used to retain iodine and aerosols radioactive releases coming from the containment atmosphere. However, during a SA, large quantities of hydrogen can also be generated. Hydrogen reacts violently with oxygen and its combustion could impair systems, components, or structures. For this reason, to protect the integrity of the FCVS against hydrogen explosions, an inertization system is found necessary. This system should create an inert atmosphere previous to any containment venting that impedes the contact of hydrogen and oxygen. In this paper, the inertization system for Cofrentes NPP is presented. It consists of a nitrogen injection located in three different points. A computational model of the FCVS as well as the inertization system has been created. The results show that if the nitrogen sweeps and the containment venting are properly synchronized, the hydrogen risk could be reduced to a minimum and therefore, the integrity of the FCVS would be preserved.

Supercritical fluids show great potential as future coolants for nuclear reactors, thermal power, and solar power plants. Compared to the subcritical condition, supercritical fluids show advantages in heat transfer due to thermodynamic properties near the critical point. A specific field of interest is an innovative decay heat removal system for nuclear power plants, which is based on a turbine-compressor system with supercritical CO 2 as the working fluid. In case of a severe accident, this system converts the decay heat into excess electricity and low-temperature waste heat, which can be emitted to the ambient air. To guarantee the retrofitting of this decay heat removal system into existing nuclear power plants, the heat exchanger (HE) needs to be as compact and efficient as possible. Therefore, a diffusion-bonded plate heat exchanger (DBHE) with mini channels was developed and manufactured. This DBHE was tested to gain data of the transferable heat power and the pressure loss. A multipurpose facility has been built at Institut für Kernenergetik und Energiesysteme (IKE) for various experimental investigations on supercritical CO 2 , which is in operation now. It consists of a closed loop where the CO 2 is compressed to supercritical state and delivered to a test section in which the experiments are run. The test facility is designed to carry out experimental investigations with CO 2 mass flows up to 0.111 kg/s, pressures up to 12 MPa, and temperatures up to 150 °C. This paper describes the development and setup of the facility as well as the first experimental investigation.

Evaluation of fuel debris properties in the Fukushima Daiichi nuclear power plant (1F) is required to develop fuel debris removal tools. In the removal of debris resulting from the Three Mile Island unit 2 (TMI-2) accident, a core-boring system played an important role. Considering the working principle of core boring, hardness, elastic modulus, and fracture toughness were found to be important fuel debris properties that profoundly influenced the performance of the boring machine. It is speculated that uranium and zirconium oxide solid solution (U,Zr)O 2 is one of the major materials in the fuel debris from 1F. In addition, the Zr content of the fuel debris from 1F is expected to be higher than that of the debris from TMI-2 because the 1F reactors were boiling-water reactors. In this research, the mechanical properties of cubic (U,Zr)O 2 samples containing 10%–65% ZrO 2 are evaluated. The hardness, elastic modulus, and fracture toughness are measured by the Vickers test, ultrasonic pulse echo method, and indentation fracture method, respectively. In the case of (U,Zr)O 2 samples containing less than 50% ZrO 2 , Vickers hardness and fracture toughness increased, and the elastic modulus decreased slightly with increasing ZrO 2 content. Moreover, all of those values of the (U,Zr)O 2 samples containing 65% ZrO 2 increased slightly compared to (U,Zr)O 2 samples containing 55% ZrO 2 . ZrO 2 content affects fracture toughness significantly in the case of samples containing less than 10% ZrO 2 . Higher Zr content (exceeding 50%) has little effect on the mechanical properties.

This work focuses on the effect of dissolved oxygen concentration in liquid lead-bismuth eutectic (LBE) on the onset of dissolution corrosion in a solution-annealed 316 L austenitic stainless steel. Specimens made of the same 316 L stainless steel heat were exposed for 1000 h at 450 °C to static liquid LBE with controlled concentrations of dissolved oxygen, i.e., 10 −5 , 10 −6 , and 10 −7 mass%. The corroded 316 L steel specimens were analyzed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). A complete absence of dissolution corrosion was observed in the steel specimens exposed to liquid LBE with 10 −5 and 10 −6 mass% oxygen. In the same specimens, isolated “islands” of FeCr-containing oxides were also detected, indicating the localized onset of oxidation corrosion under these exposure conditions. On the other hand, dissolution corrosion with a maximum depth of 59 μ m was detected in the steel specimen exposed to liquid LBE with 10 −7 mass% oxygen. This suggests that the threshold oxygen concentration associated with the onset of dissolution corrosion in this 316 L steel heat lies between 10 −6 and 10 −7 mass% oxygen for the specific exposure conditions (i.e., 1000 h, 450 °C, static liquid LBE).

In a natural circulation boiling water reactor (BWR), the core power varies in both axial and radial directions inside the reactor core. The variation along the axial direction is more or less constant throughout the reactor; however, there exists variation of reactor power in the radial direction. The channels located at the periphery have low power compared to the center of the core and are equipped with orifices at their inlet. This creates nonuniformity in the radial direction in the core. This study has been performed in order to understand the effect of this radial variation of power on the stability characteristics of the reactor. Four channels of a pressure tube type natural circulation BWR have been considered. The reactor has been modeled using RELAP5/MOD 3.2. Before using the model, it was first benchmarked with experimental measurements and then the characteristics of both low power and high power oscillations, respectively, known as type-I and type-II instability, have been investigated. It was observed that the type-I instability shows slight destabilizing effect of increase in power variation among different channels. However, in the case of type-II instability, it was found out that the oscillations get damped with an increase in power variation among the channels. A similar effect was found for the presence of orifices at the inlet in different channels. However, the increase in number of orificed channels showed stabilizing effect for both type-I and type-II instabilities.

The paper presents the computational fluid dynamics (CFD) combustion modeling approach based on two combustion models. This modeling approach was applied to a hydrogen deflagration experiment conducted in a large-scale confined experimental vessel. The used combustion models were Zimont's turbulent flame-speed closure (TFC) model and Lipatnikov's flame-speed closure (FSC) model. The conducted simulations are aimed to aid identifying and evaluating the potential hydrogen risks in nuclear power plant (NPP) containment. The simulation results show good agreement with experiment for axial flame propagation using the Lipatnikov combustion model. However, substantial overprediction in radial flame propagation is observed using both combustion models, which consequently results also in overprediction of the pressure increase rate and overall combustion energy output. As assumed for a large-scale experiment without any turbulence inducing structures, the combustion took place in low-turbulence regimes, where the Lipatnikov combustion model, due to its inclusion of quasi-laminar source term, has advantage over the Zimont model.

The understanding of the radial distribution of temperature in a fuel pellet, under normal operation and accident conditions, is important for a safe operation of a nuclear reactor. Therefore, in this study, we have solved the steady-state heat conduction equation, to analyze the temperature profiles of a 12 mm diameter cylindrical dispersed nuclear fuels of U 3 O 8 -Al, U 3 Si 2 -Al, and UN-Al operating at 597 ° C. Moreover, we have also derived the thermal conductivity correlations as a function of temperature for U 3 Si 2 , uranium mononitride (UN), and Al. To evaluate the thermal conductivity correlations of U 3 Si 2 , UN, and Al, we have used density functional theory (DFT) as incorporated in the Quantum ESPRESSO (QE) along with other codes such as Phonopy, ShengBTE, EPW (electron-phonon coupling adopting Wannier functions), and BoltzTraP (Boltzmann transport properties). However, for U 3 O 8 , we utilized the thermal conductivity correlation proposed by Pillai et al. Furthermore, the effective thermal conductivity of dispersed fuels with 5, 10, 15, 30, and 50 vol %, respectively of dispersed fuel particle densities over the temperature range of 27–627 °C was evaluated by Bruggman model. Additionally, the temperature profiles and temperature gradient profiles of the dispersed fuels were evaluated by solving the steady-state heat conduction equation by using Maple code. This study not only predicts a reduction in the centerline temperature and temperature gradient in dispersed fuels but also reveals the maximum concentration of fissile material (U 3 O 8 , U 3 Si 2 , and UN) that can be incorporated in the Al matrix without the centerline melting. Furthermore, these predictions enable the experimental scientists in selecting an appropriate dispersion fuel with a lower risk of fuel melting and fuel cracking.

Studies on debris bed formation behavior are important for improved evaluation of core relocation and debris bed coolability that might be encountered in a core disruptive accident (CDA) of sodium-cooled fast reactors (SFR). Motivated to clarify the flow-regime characteristics underlying this behavior, both experimental investigations and empirical-model development are being performed at the Sun Yat-sen University in China. As for the experimental study, several series of simulated experiments are being conducted by discharging various solid particles into water pools. To obtain a comprehensive understanding, a variety of experimental parameters, including particle size (0.000125– 0.008 m), particle density (glass, aluminum, alumina, zirconia, steel, copper, and lead), particle shape (spherical and nonspherical), and water depth (0–0.8 m) along with the particle release pipe diameter (0.01–0.04 m) were varied. It is found that due to the different interaction mechanisms between solid particles and water pool, four kinds of flow regimes, termed, respectively, as the particle-suspension regime, the pool-convection dominant regime, the transitional regime, and the particle-inertia dominant regime, were identifiable. As for the empirical-model development, aside from a base model which is restricted to predictions of spherical particles, in this paper considerations on how to cover more realistic conditions (esp. debris of nonspherical shapes) are also discussed. It is shown that by coupling the base model with an extension scheme, respectable agreement between experiments and model predictions for regime transition can be achieved for both spherical and nonspherical particles given our current range of conditions.

The Canadian super critical water-cooled reactor (SCWR) concept requires materials to operate at higher temperatures than current generation III water-cooled reactors. Materials performance after radiation damage is an important design consideration. Materials that are both corrosion resistant and radiation damage tolerant are required. This paper summarizes the operating conditions including temperature, neutron flux, and residence time of in-core Canadian SCWR components. The focus is on the effects of irradiation on in-core components, including those exposed to a high neutron flux in the fuel assembly, the high pressure boundary between coolant and moderator, as well as the low-temperature, low-flux calandria vessel that contains the moderator. Although the extreme conditions and the broad range of SCWR in-core operating conditions present significant materials selection challenges, candidate alloys that can meet the performance requirements under most in-core conditions have been identified. However, for all candidate materials, insufficient data are available to unequivocally ensure acceptable performance and experimental irradiations of candidate core materials will be required. Research programs are to include out-of-pile tests on un-irradiated and irradiated alloys. Ideally, in-flux studies at appropriate temperatures, neutron spectrum, dose rate, duration, and coolant chemistry will be required. Characterization of the microstructure and the mechanical behavior including strength, ductility, swelling, fracture toughness, cracking, and creep on each of the in-core candidate materials will ensure their viability in the Canadian SCWR.

Most of the dry storage cask systems that are used to contain nuclear fuel are austenitic stainless steel canisters. Past experience suggests that stainless steel is susceptible to stress corrosion cracking (SCC) in the presence of chloride salts. A crack growth rate (CGR) model has been developed and applied to evaluate the crack depth in stainless steel canisters over the timeframe of the storage at independent spent fuel storage installations. This study focuses on stainless steel canisters for dry storage systems in the two nuclear power plants in Taiwan. The crack depth was first evaluated using the CGR model, site climate data, and canister surface temperature. The critical crack size and depth were then determined from the structural tolerance assessment of the canister shells. It was found that the variations in the thermal and hydraulic properties of dry storage canisters produce large variations in the SCC initiation time but do not affect the surface temperature in the range of 55–60 °C. The CGR at the SCC initiation is high and the growth of flaws is significant. The surface temperature and CGR decrease with time. The total crack depth therefore may not vary greatly as a function of SCC initiation time. Overall, dry storage canisters show high structural tolerance to crack size and depth.

In recent years, the morphological characteristics and stabilization methods of free interface in liquid windowless target become hot research topics in accelerator driven subcritical system (ADS). Based on the structure design of a certain windowless spallation target, computational fluid dynamics (CFD) software of CFX was used to simulate and analyze its free interface character. The method of k – ε turbulence, cavitation, and volume of fluid (VOF) model was used to study the flow characteristic of liquid Lead-Bismuth eutectic (LBE) alloy with cavitation phase change and to analyze the free interface morphology characteristics of coolant in the target area. It is concluded that the target region forms two stable free interfaces when fluid outlet pressure is in the range of 10–40 kPa and fluid entrance velocity is in the range of 0.5–1.2 m/s. The flow field near the free interface structure is complex. The vortex region appears, and the disorders in the vortex flow pattern lead to fluctuation of the free interface. After the study of stable free interface morphology establishing process, heat transfer characteristic of windowless target was further analyzed.

The transient reactor test facility (TREAT), a graphite moderated experimental reactor, is scheduled to restart in late 2017. There is now renewed interest in development of capabilities to model and simulate the TREAT transients using three-dimensional coupled physics. To validate existing transient analysis tools as well as those under development, several temperature-limited transients have been modeled and analyzed. These transients are from the M8 calibration (M8CAL) experiment series, a set of experiments performed to calibrate the reactor detectors for the planned M8 series of fuel tests. Detailed reactor models were prepared that were then used to calculate the pretransient and post-transient k eff values as well as corresponding reactivity insertions. Alterations to modeled values of shutdown and initial transient rod insertion depths were made to better match the reported experimental values of reactivity insertions assuming just critical pretransient states. It was found that two of the altered media inputs, fuel and Zircaloy-3 cladding, had significant effects on the k eff . In addition, increasing shutdown rod insertion by 3–5 cm and decreasing initial transient rod insertion by 1–2 cm gave perfect pretransient k eff and total reactivity insertion values. However, the revised positions are as much as a factor of 3–20 different from reported uncertainty of 0.762 cm. This suggests that boron concentration uncertainties may play a significant role in accurately modeling the TREAT transients and should be investigated thoroughly.

The intermittency of renewable power generation systems on the low carbon electric grid can be alleviated by using nuclear systems as quasi-storage systems. Nuclear air-Brayton systems can produce and store hydrogen when electric generation is abundant and then burn the hydrogen by co-firing when generation is limited. The rated output of a nuclear plant can be significantly augmented by co-firing. The incremental efficiency of hydrogen to electricity can far exceed that of hydrogen in a standalone gas turbine. Herein, we simulate and evaluate this idea on a 50 MW small modular liquid metal/molten salt reactor. Considerable power increases are predicted for nuclear air-Brayton systems by co-firing with hydrogen before the power turbine.