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ASME J of Nuclear Rad Sci. 2018;4(3):031001-031001-8. doi:10.1115/1.4039438.

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, fi, 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 fi 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 fi correlation with the viscosity ratio and the improved fi 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).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031002-031002-24. doi:10.1115/1.4039594.

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/MOD3.2. 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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031003-031003-13. doi:10.1115/1.4039597.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031004-031004-11. doi:10.1115/1.4039595.

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 CO2 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 CO2, which is in operation now. It consists of a closed loop where the CO2 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 CO2 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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031005-031005-10. doi:10.1115/1.4038367.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031006-031006-10. doi:10.1115/1.4039035.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031007-031007-14. doi:10.1115/1.4039343.

The Great East Japan Earthquake on Mar. 11, 2011 triggered huge tsunami waves that attacked Fukushima Daiichi Nuclear Power Plant (Fukushima-1). Units 1, 3, and 4 had hydrogen explosions. Units 1–3 had core meltdowns and released a large amount of radioactive material. Published investigation reports did not explain how the severity of the accident could have been prevented. We formed a study group to find: (A) Was the earthquake-induced huge tsunami predictable at Fukushima-1? (B) If it was predictable, what preparations at Fukushima-1 could have avoided the severity of the accident? Our conclusions were: (a) The tsunami that hit Fukushima-1 was predictable, and (b) the severity could have been avoided if the plant had prepared a set of equipment, and most of all, had exercised actions to take against such tsunami. Necessary preparation included: (1) a number of direct current (DC) batteries, (2) portable underwater pumps, (3) portable alternating current (AC) generators with sufficient gasoline supply, (4) high voltage AC power trucks, and (5) drills against extended loss of all electric power and seawater pumps. This set applied only to this specific accident. A thorough preparation would have added (6) portable compressors, (7) watertight modification to reactor core isolation cooling system (RCIC) and high pressure coolant injection system (HPCI) control and instrumentation, and (8) fire engines for alternate low pressure water injection. Item (5), i.e., to study plans and carry out exercises against the tsunami would have identified all other necessary preparations.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031008-031008-8. doi:10.1115/1.4039848.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031009-031009-6. doi:10.1115/1.4039067.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031010-031010-9. doi:10.1115/1.4039036.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031011-031011-7. doi:10.1115/1.4039847.

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)O2 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)O2 samples containing 10%–65% ZrO2 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)O2 samples containing less than 50% ZrO2, Vickers hardness and fracture toughness increased, and the elastic modulus decreased slightly with increasing ZrO2 content. Moreover, all of those values of the (U,Zr)O2 samples containing 65% ZrO2 increased slightly compared to (U,Zr)O2 samples containing 55% ZrO2. ZrO2 content affects fracture toughness significantly in the case of samples containing less than 10% ZrO2. Higher Zr content (exceeding 50%) has little effect on the mechanical properties.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031012-031012-6. doi:10.1115/1.4039600.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031013-031013-11. doi:10.1115/1.4039066.

One of the key elements in probabilistic risk assessment is the identification and characterization of uncertainties. This paper suggests a procedure to identify influencing factors for uncertainty in source term evaluation, which are important to risk of public dose. We propose the following six steps for the identification in a systematic manner in terms of completeness and transparency of the results using both a logic diagram based on basic equations and expert opinions: (1) identification of uncertainty factors based on engineering knowledge of accident scenario analysis; (2) derivation of factors at the level of physical phenomena and variable parameters by expansion of dynamic equation for the system and scenario to be investigated, (3) extraction of uncertainties in variable parameters; (4) selection of important factors based on sensitivity study results and engineering knowledge; (5) identification of important factors for uncertainty analysis using expert opinions; and (6) integration of selected factors in the aforementioned steps. The proposed approach is tested with a case study for a risk-dominant accident scenario in direct cycle high-temperature gas-cooled reactor (HTGR) plant. We use this approach for evaluating the fuel temperature in terms of reactor dynamics and thermal hydraulic characteristics during a depressurized loss-of-forced circulation (DLOFC) accident with the failure of mitigation systems such as control rod systems (CRS) in a representative HTGR plan. In total, six important factors and 16 influencing factors were successfully identified by the proposed method in the case study. The selected influencing factors can be used as input parameters in uncertainty propagation analysis.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031014-031014-11. doi:10.1115/1.4039636.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031015-031015-9. doi:10.1115/1.4039037.

A fast breeder reactor (FBR) is considered as the promising technology in terms of load reduction on the environment, because the FBR has capability to improve usage efficiency of uranium resources and can reduce high-level radioactive waste which needs to be managed for millions of years. A cold trap is one of the important components in the FBR to control the impurity concentration of the liquid sodium. For accurate evaluation of the cold trap performance, we have been proposing the three-dimensional (3D) numerical analysis method of the cold trap. In this method, the evaluation of the impurity precipitation phenomena on the surface of the mesh wire of the cold trap is the key. For this, the numerical analysis method which is based on the lattice kinetic scheme (LKS) has been proposed. In order to apply the LKS to the impurity precipitation simulation of the cold trap, two models (the low Reynolds number model and the impurity precipitation model) have been developed. In this paper, we focused on the validation of these models. To confirm the validity of the low Reynolds number model, the Chapman–Enskog analysis was applied to the low Reynolds number model. As a result, it has been theoretically confirmed that the low Reynolds number model can recover the correct macroscopic equations (incompressible Navier–Stokes equations) with small error. The low Reynolds number model was also validated by the numerical simulation of two-dimensional (2D) channel flow problem with the low Reynolds number conditions which correspond to the actual cold trap conditions. These results have confirmed that the error of the low Reynolds number model is ten times smaller than that of the original LKS. The validity of the impurity precipitation model was investigated by the comparison to the precipitation experiments. In this comparison, the mesh convergence study was also conducted. These results have confirmed that the proposed impurity precipitation model can simulate the impurity precipitation phenomena on the surface of the mesh wire. It has been also confirmed that the proposed impurity precipitation model can simulate the impurity precipitation phenomenon regardless of the cell size which were tested in this investigation.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031016-031016-13. doi:10.1115/1.4038595.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031017-031017-7. doi:10.1115/1.4038929.

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 keff 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 keff. In addition, increasing shutdown rod insertion by 3–5 cm and decreasing initial transient rod insertion by 1–2 cm gave perfect pretransient keff 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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031018-031018-5. doi:10.1115/1.4039596.

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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031019-031019-7. doi:10.1115/1.4039598.

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 μm 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).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031020-031020-7. doi:10.1115/1.4039886.

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 U3O8-Al, U3Si2-Al, and UN-Al operating at 597 °C. Moreover, we have also derived the thermal conductivity correlations as a function of temperature for U3Si2, uranium mononitride (UN), and Al. To evaluate the thermal conductivity correlations of U3Si2, 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 U3O8, 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 (U3O8, U3Si2, 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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031021-031021-5. doi:10.1115/1.4039034.

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 kε 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.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):031022-031022-7. doi:10.1115/1.4039501.

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.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2018;4(3):034501-034501-4. doi:10.1115/1.4039694.

During the process of beam extraction in positive ion source under high voltage region, a large number of electrons are produced in the gaps of grids. After back-streaming acceleration, these electrons go back to arc chamber or impinge grids and then heat back plate or grids, which are harmful for the safety of ion source. Under the situation of poor beam extraction optics, a large part of the primary beam ions bombard the surface of suppressor grid (SG). And this process produces a large number of electrons. Due to the huge extracted voltage, the secondary electron emission coefficient of the SG surface is also high. As a result, the grids' current grows. According to the measurement of the current of SG and the calculation of the perveance of the corresponding shoot, the effect of ion beam divergence angle on back-streaming electrons can be analyzed. When the beam divergence angle increases, the number of back-streaming electrons increases rapidly, and grids' current changes significantly, especially the current of gradient grid and SG. The results can guide the parameters operating on the ion source for Experimental Advanced Superconducting Tokamak-neutral beam injection (EAST-NBI) and find the reasonable operation interval of perveance and to ensure the safety and stable running of the ion source, which has great significance on the development of long pulse, high power ion source.

Topics: Electrons , Ions , Silencers , Heat
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):034502-034502-3. doi:10.1115/1.4039599.

From 1953 to now, it has happened six times so-called red oil explosion accidents worldwide, resulting in different degrees of equipment and construction damage and environmental contamination. And research related to red oil has never stopped. Preventive measures for red oil explosion were established in some reports, and these measures provided good practice experience and reference for other countries. Nevertheless, research conclusions and knowledge of red oil vary from country to country. Especially, investigations on stability of tributyl phosphate (TBP)—nitric system was made in recent years, and the results indicated that the red oil runway reaction will happen even in lower temperature and lower nitric acid concentration in contrast with the reported value. Therefore, in order to facilitate future study on red oil explosion, related research results of red oil explosion accidents were combed in this paper, and the characters of study work of different periods were also summarized, and definition, formation conditions of red oil were analyzed and compared, as well as the new viewpoints of recent literatures.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(3):034503-034503-5. doi:10.1115/1.4039885.

Uranium dioxide (UO2) is the typical fuel that is used in the current nuclear power plant; fission gas atoms are produced during and after the nuclear reactor operation; the fission gas atoms have a significant effect on the performance of UO2 fuel in the nuclear reactor. In this paper, we investigated the diffusion of the fission gas atoms in the UO2 fuel by using the first-principles calculation method based on the density functional theory (DFT). The results indicate that the volume of the UO2 cell increased when there is a fission gas atom enters in the UO2 supercell; the elastic properties of UO2 are in good agreement with other simulation results and experimental data and the fission gas atoms make the ductility of UO2 decreased; the fission gas atoms prefer to occupy the octahedral interstitial site (OIS) over the uranium vacancy site and the oxygen vacancy site, and the oxygen vacancy site is the most difficult occupied site due to the formation of an oxygen vacancy is more difficult than that of the uranium vacancy; the diffusion barrier of a He atom in the UO2 supercell is higher than that of an oxygen atom, that means that the diffusion of the He atom in UO2 fuel is weaker than that of the oxygen atom. Our works may shed some light on the formation mechanism of the bubbles caused by the fission gas atoms in the UO2 fuel.

Commentary by Dr. Valentin Fuster

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