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Guest Editorial

Commentary by Dr. Valentin Fuster

SPECIAL SECTION: SELECTED PAPERS FROM THE INTERNATIONAL YOUTH NUCLEAR CONGRESS 2018 - 26TH WIN GLOBAL ANNUAL CONFERENCE

ASME J of Nuclear Rad Sci. 2019;5(2):020901-020901-8. doi:10.1115/1.4042193.

Treatment and conditioning of spent ion exchange resins from nuclear facilities is a complex process that not only should contemplate obtaining a stable product suitable for long-term storage and/or disposal, but also have to take into account the treatment of secondary currents generated during the process. The combination of low temperature pyrolysis treatment and high performance plasma treatment (HPPT) of the off-gas generated could be a novel solution for organic matrix nuclear wastes with economic and safety advantages. In the present work, results of lab scale studies associated with the pyrolysis off-gas characterization and the performance and operating parameters influence on the removal of model compounds in a laboratory-scale flow reactor, using inductively coupled plasma under subatmospheric conditions, are shown. The pyrolysis off-gas stream was largely characterized and the evolution of main compounds of interest as function of temperature process was established. The results of plasma assays with the model compound demonstrate a high destruction and removal efficiency (>99.990%) and a good control over the final gas products. First results of a bench scale arrangement combining both processes are presented and bode well for the application of this combined technology.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020902-020902-8. doi:10.1115/1.4042363.

Difficulties are experienced during the thermal–hydraulic design of a nuclear reactor operating in the transition flow regime and are the result of the inability to accurately predict the heat transfer coefficient (HTC). Experimental values for the HTC in rectangular channels are compared with the calculated by correlations usually used for the design of material testing reactors (MTR). The values predicted by Gnielinski and Kreith correlations at Reynolds numbers below 5000 are not necessarily conservative. The Al-Arabi-Churchill correlation with the correction proposed by Jones has proved to be conservative for Reynolds between 2100 and 5000. Two alternative design approaches are proposed to solve a specific thermal–hydraulic design problem for a MTR operating at Reynolds 2500. The conservative approach comprises two alternatives: the use of Al-Arabi correlation with no uncertainty factors, as it has proved to be conservative, or the use of Kreith correlation with a maximum uncertainty. In this conservative approach, maximum deviations in other input parameters are also taken into account. The best estimate plus uncertainty approach considers an uncertainty distribution in input parameters to generate a random sample of 59 inputs. An uncertainty distribution based on the ratio between the experimental and the calculated HTC, when using Kreith correlation, is considered. Results are given in terms of maximum and minimum bounds for the figure of merit used as design criterion with 95% probability and 95% confidence level. The best estimate plus uncertainty approach offers a less penalizing design and its use depends on regulator's acceptance.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020903-020903-11. doi:10.1115/1.4042705.

DIONISIO is a computer code designed to simulate the behavior of one nuclear fuel rod during its permanence within the reactor. Starting from the power history and the external conditions to which the rod is subjected, the code predicts all the meaningful variables of the system. Its application range has been recently extended to include accidental conditions, in particular the so-called loss of coolant accidents (LOCA). In order to make realistic predictions, the conditions in the rod environment have been taken into account since they represent the boundary conditions with which the differential equations describing the fuel phenomena are solved. Without going into the details of the thermal-hydraulic modeling, which is the task of the specific codes, a simplified description of the conditions in the cooling channel during a LOCA event has been developed and incorporated as a subroutine of DIONISIO. This has led to an improvement of the fuel behavior simulation, which is evidenced by the considerable number of comparisons with experiments carried out, many of them reported in this paper. Moreover, this work describes a model of high temperature capture and release of hydrogen in the nuclear fuel cladding, in scenarios typical of LOCA events. The corresponding computational model is being separately tested and will be next included in the DIONISIO thermal-hydraulic module.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020904-020904-8. doi:10.1115/1.4042118.

During operation of light water reactors, the Zircaloy fuel rod cladding is susceptible for hydrogen uptake. When the local solubility limit of hydrogen in Zircaloy is reached, additional hydrogen precipitates as zirconium hydride, which affects the ductility of the fuel rod cladding. Especially, the radially aligned hydrides enhance embrittlement, while circumferential (azimuthal) hydrides have a less detrimental effect. In this work, the influence of high temperatures during the dry storage period on hydride dissolution and precipitation is demonstrated. Therefore, in a descriptive example scenario being discussed, the simulation of a limited heat removal from the cask will heat up the dry storage cask for days and causes dissolution of hydrides in the cladding. Depending on the threshold stress for reorientation, the following cooldown results on different hydride precipitation behavior. The threshold stress leads to an enhanced or delayed precipitation of radial hydrides. The GRS fuel rod code TESPA-ROD is equipped with a new model for hydrogen solubility and applied to long-term storage transients. In this article, hydride refers to zirconium hydrides formed inside the fuel rod cladding.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020905-020905-7. doi:10.1115/1.4042360.

The study presented in this paper is part of the technological surveillance performed at the Electricité De France (EDF) Research and Development (R&D) Center, in the Pericles department, and investigates the feasibility of modeling in-vessel melt retention (IVMR) phenomena for small modular reactors (SMR) with the modular accident analysis program version 5 in its EDF proprietary version (MAAP5_EDF), applying conservative hypotheses, such as constant decay heat after corium relocation to the lower head. The study takes advantage of a corium stratification model in the lower head of the vessel, developed by EDF R&D for large-sized prospective pressurized water reactors (PWRs). The analysis is based on a stepwise approach in order to evaluate the impact of various effects during IVMR conditions. First, an analytical calculation is performed in order to establish a reference case to which the MAAP5_EDF code results are compared. In a second step, the impact of the lower head geometry, vessel steel ablation, and subsequent relocation on the heat flux has been analyzed for cases where heat dissipation through radiation is neglected (in first approximation). Finally, the impact of heat losses through radiation as well as the crust formation around the pool has been assessed. The results demonstrate the applicability of the MAAP5_EDF code to SMRs, with heat fluxes lower than 1.1 MW/m2 for relevant cases, and identify modeling improvements.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020906-020906-7. doi:10.1115/1.4042192.

Nuclear energy has created controversy since its conception during the 1950s. Arguments against it have been constant through the years until the current state on which the majority of western societies are against it as seen in recent surveys. Additionally, confidence in science and scientists is also relatively low. In Spain, these two facts are related with science alphabetization; an average person with lower science alphabetization tends to be more negative about science, and specifically, about nuclear energy science. In this aspect, as science affects major decisions in society, in democracy, it is important that the public is able to interpret scientific information. It is in this context that Jóvenes Nucleares appeared: an organization created by the Spanish Nuclear Society and formed by young people interested in nuclear energy. One of its main goals was the spreading of nuclear science into society. This was made through lectures at high schools, content creation, and enveloping communication campaigns. A skeptical approach has always been taken trying to separate from the lobby argumentation and promoting a strong critical thought. In this paper, as an example of communication campaign, the Basic Nuclear Fusion Course is presented. This campaign involved the creation of the informative content, gathering it into a book, the development of a lecture (consisting of nine topics related to nuclear fusion) to be delivered in universities or high schools, and a strong advertisement effort through social media and presentations in congresses. This campaign has been possible thanks to voluntary work; the main cost of the campaign was the book printing. The early results predict a great support to this new format included into the Jóvenes Nucleares divulgation activities as perceived in the attendance and feedback provided by the audience. With these activities, Jóvenes Nucleares aspires to put another grain of sand toward narrowing the gap between science and society.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020907-020907-6. doi:10.1115/1.4042497.

Glasses have emerged as alternative materials that can be used for long-term treatment and management of radioactive waste. Specifically, glasses can be used as a matrix to immobilize the radioactive material. Within the glass industry, silicate glasses are the most widely used due to their properties and to the large knowledge existent about them. Alkaline free silicate glasses are particularly corrosion resistant. Due to the latter, rare earth aluminosilicate glasses are good candidates for actinides immobilization, especially, yttrium aluminosilicate (YAS) glasses. The crystallization kinetics of YAS glasses on heating has been already studied, and this work is focused on the effect of lutetium addition on the YAS glass crystallization kinetics. The presence of a small amount of lutetium in a YAS glass decreases the surface density of nucleation sites (Ns) by about 1 order of magnitude and significantly decreases the crystal growth rate (U). In this work, it was observed that lutetium additions on the order of 0.2 (wt %) to a YAS glass dramatically decreased Ns, for example, at 1000 °C from 1011 to 109 nuclei/m2. Additionally, U for yttrium disilicate phase decreased from (8.21 ± 0.28) μm/h to (0.54 ± 0.04) μm/h at the same temperature.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020908-020908-9. doi:10.1115/1.4042119.

In the event of a loss of integrity of the main coolant line, a large mass and energy release from the primary circuit to the containment is to be expected. The temporal evolution of such depressurization is mainly governed by the critical flow, whose correct prediction requires, in first place, a correct description of the different friction terms. Within this work, selected friction models of the CESAR module of the Accident Source Term Evaluation Code (ASTEC) V2.1 integral code are validated against data from the Moby Dick test facility. Simulations are launched using two different numerical schemes: on the one hand, the classical five equation (drift flux) approach, with one momentum conservation equation for an average fluid plus one algebraic equation on the drift between the gas and the liquid; on the other hand, the recently implemented six equation approach, where two differential equations are used to obtain the phase velocities. The main findings are listed hereafter: The use of five equations provides an adequate description of the pressure loss as long as the mass fluxes remain below 1.24 kg/cm2 s and the gas mass fractions below 5.93 × 10 − 4. Beyond those conditions, the hypotheses of the drift flux model are exceeded and the use of an additional momentum equation is required. The use of an additional momentum equation leads to a better agreement with the experimental data for a wider range of mass fluxes and gas mass fractions. However, the qualitative prediction for high gas mass fractions still shows some deviations due to the decrease of the regular friction term at the end of the test section.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020908-020908-9. doi:10.1115/1.4042119.

In the event of a loss of integrity of the main coolant line, a large mass and energy release from the primary circuit to the containment is to be expected. The temporal evolution of such depressurization is mainly governed by the critical flow, whose correct prediction requires, in first place, a correct description of the different friction terms. Within this work, selected friction models of the CESAR module of the Accident Source Term Evaluation Code (ASTEC) V2.1 integral code are validated against data from the Moby Dick test facility. Simulations are launched using two different numerical schemes: on the one hand, the classical five equation (drift flux) approach, with one momentum conservation equation for an average fluid plus one algebraic equation on the drift between the gas and the liquid; on the other hand, the recently implemented six equation approach, where two differential equations are used to obtain the phase velocities. The main findings are listed hereafter: The use of five equations provides an adequate description of the pressure loss as long as the mass fluxes remain below 1.24 kg/cm2 s and the gas mass fractions below 5.93 × 10 − 4. Beyond those conditions, the hypotheses of the drift flux model are exceeded and the use of an additional momentum equation is required. The use of an additional momentum equation leads to a better agreement with the experimental data for a wider range of mass fluxes and gas mass fractions. However, the qualitative prediction for high gas mass fractions still shows some deviations due to the decrease of the regular friction term at the end of the test section.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020909-020909-7. doi:10.1115/1.4042358.

One of the limiting conditions during operation of a pressurized water reactor (PWR) is cladding integrity in class I (normal operations) or class II (most frequent). Cladding integrity is limited typically by the departure from the nucleate boiling (DNB), which criterion ensures an appropriate core cooling. Adequate heat transfer between the fuel cladding and reactor coolant is achieved by preventing DNB that is avoided if the local heat flux is lower than the critical heat flux (CHF). The DNB is estimated thanks to thermal-hydraulic (TH) design codes, as the VIPRE-W code that predicts the fluid behavior based on the geometry of the problem, the fuel rods and the fluid properties among others. One of the parameters that influences the DNB estimation is the thermal diffusion coefficient (TDC), which depends on the fuel design and is affected by the grid spacing. As a matter of fact, the TDC enters into the DNB calculation for thermal mixing between subchannels and in some special cases like the most primitive fuel designs, as a factor within the DNB correlation. Nevertheless, although the TDC is a variable, the TH design codes used for the DNB prediction consider the TDC as a constant. This investigation is founded on a new numerical program developed to explore the effect of the TDC on the DNB. In addition to this, variables as the effect of the turbulent momentum factor (FTM) and the correlation effect has been explored too. The most direct outcome of this research is the substantial extension of the existing studies of VIPRE-W TH code. The results show that TDC has an effect on the DNB dominated by the radial power distribution. The departure from nucleate boiling ratio (DNBR) increases up to 1.2% when TDC is a variable under normal operation radial shapes. For the design radial distribution, this effect is vanished but observable for values under 0.02 with an exponential increase of the DNBR with respect to the TDC. From this moment on, the energy exchanged between subchannels is negligible due to the flatness shape of the radial enthalpy distribution.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020910-020910-5. doi:10.1115/1.4042362.

Uranium tetrafluoride was synthesized using a novel method, which consists of a combination of carbochlorination reaction between a mixture of U3O8 and sucrose carbon with chlorine, and a solid-state halogen exchange reaction between the products of the carbochlorination reaction and sodium fluoride. The thermodynamic feasibility to produce the halogen exchange reaction between UCl4 and NaF was analyzed. Reactions are favorable in standard conditions, even at low temperature. We have prepared a mixture of UCl4 and UCl2O2 by U3O8 treatment in Cl2 atmosphere with presence of sucrose carbon at 900 °C. UCl4 and UCl2O2 were obtained as a condensed product, which was collected in a quartz capsule containing NaF. The capsules were sealed after several repeated stages of argon purges and mechanic vacuum. Subsequently, they were treated at 300–350 °C for 2 h. We obtained that when NaF is the limiting reagent, the solid product of the thermal treatment of the capsules consists in a mixture of UF4 and NaCl. Solid products were characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM), fluorescence spectroscopy, and energy dispersive X-ray spectroscopy. Gaseous products were identified by Fourier-transform infrared spectroscopy.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020911-020911-9. doi:10.1115/1.4042498.

The American Society of Mechanical Engineers (ASME) Section III Rules for Construction of Nuclear Facility Components subsection NB-2331 Material for Vessels requires that the effects of irradiation shall be considered on material toughness properties in the core belt line region of the reactor vessel. The code also states that “the design specifications shall include additional requirements, as necessary, to ensure adequate fracture toughness for the service lifetime of the vessel.” In a design report of nuclear pressure vessel, the design and service loads do not include loads that are affected by fracture toughness of the material. However, in the cases of fitness-for-service assessment for component flaws (prevalent with age of component), irradiated material properties become highly relevant. An example of a fitness-for-service is that of a beyond design basis reactor vessel head drop accident in a pressurized water reactor with a nozzle junction flaw. As a case study, the critical size of a postulated external surface semielliptical circumferential crack in the combustion engineering three-loop pressurized water reactor nozzle–vessel junction is calculated using ansys Workbench (Academic version) with the applied impact load from the vessel head drop accident. Failure assessment diagrams for numerous crack depths and lengths were developed considering the fracture toughness properties of the irradiated reactor vessel steel. The mode I stress intensity results used in the failure assessment diagram were compared with the available finite element and the American Petroleum Institute (API) standard API 579 analytical solutions for validation, showing good agreement. From this case study, it is demonstrated that the effects of irradiation on fracture toughness become prominent at the same postulated crack size in the nozzle–vessel junction dispositioned as “safe” becomes “unsafe” in the fracture assessment diagram. Using the unified curve method, the irradiated fracture toughness data in design specification can be supplied so that it may be used in fitness-for-service analysis to account for component aging.

Commentary by Dr. Valentin Fuster
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020913-020913-10. doi:10.1115/1.4042906.

The International Atomic Energy Agency's (IAEA) Incident and Emergency Centre (IEC) has custom designed software tools to support assessment and prognosis of nuclear and radiological emergency scenarios, aimed at ensuring consistent and concise technical reports for emergency assessments. In this paper the functionality, updating and structural development of emergency communications tools is presented, that lead the user through a series of questions with the aid of instructions that will collect relevant technical details and organize them into standardized reports. These reports can be exported for use in internal communication or communication with external stakeholders. This paper discusses enhancements in the suite of tools, specifically the reactor assessment tool (RAT), which was updated, the emergency response action, and the radiological source assessment tools, which were expanded and finally the development of two dose assessment tools (DAT) for internal and external exposure to radioactive substances.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020914-020914-7. doi:10.1115/1.4042499.

The thermal stability of N, N-dimethylhydroxylamine (DMHAN) in the HNO3 solution was studied using microcalorimeter. The influence of concentration of HNO3, DMHAN, methylhydrazine (MH), atmosphere (air and nitrogen), and metals was investigated. The kinetic parameters and self-accelerating decomposition temperature (SADT) of the feed in process (stripping reagents 1BX, scouring agent 2DS, stripping reagents 2BX, and waste aqueous phase 2DW) were calculated by Advanced Kinetics and Technology Solutions (akts) thermokinetics software. The molar enthalpy of the reaction of NaNO2 with DMHAN and MH was also determined. The results show that the initial reaction temperature (T0) of DMHAN/HNO3 (HNO3: 1.5–3.0 mol/L, DMHAN: 0.05–0.8 mol/L) is increased as the acidity is reduced or the concentration of DMHAN is increased. Holding reductant MH made the induction period of the autocatalytic reaction longer. The air, nitrogen atmosphere, Fe, and the fission products (Zr, Ru) do not affect the decomposition of DMHAN, but the stainless steel made the T0 of DMHAN/HNO3 become lower. The SADT of 1BX/2DS, 2BX, and 2DW is 56 °C, 52 °C and 47 °C, respectively. The molar enthalpies of formation of the reaction of NaNO2 with DMHAN and MH are −411.3 kJ/mol, −246.0 kJ/mol, respectively.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):020915-020915-8. doi:10.1115/1.4042361.

This work shows the introduction of the Electrical Power System Analysis (etap) software as a calculation and analysis tool for power electrical systems of the nuclear power plants (NPP) under the orbit of Nucleoeléctrica Argentina S.A (NASA). Through the use of the software, the model of the electrical power system of the Atucha II NPP was developed. To test the functionality of the modeled electrical power circuit, studies of load flow and short-circuit analysis were conducted, yielding satisfactory results, which were contrasted with the plant design values. Once the model has been validated, this will be the basis for carrying out different studies in the plant through simulation. Furthermore, with the incorporation of etap as a fundamental calculation and analysis tool for power electrical systems at the company's engineering departments, it is expected to improve the safety, operation, quality, reliability, and maintenance of both the Atucha II NPP electrical power system and the other nuclear power plants operated by Nucleoeléctrica Argentina S.A.

Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2019;5(2):021001-021001-12. doi:10.1115/1.4042191.

In the design study of advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) called FA with inner duct structure (FAIDUS) is expected to enhance reactor safety during a core-disruptive accident. Evaluating the thermal-hydraulics in FAIDUS under various operating conditions is required for its design. This study is the first step toward confirming the design feasibility of FAIDUS; the thermal-hydraulics in FAIDUS are investigated with an in-house subchannel analysis code called asymmetrical flow in reactor elements (ASFRE), which can be applied to a wire-wrapped fuel pin bundle in conjunction with the distributed resistance model (DRM) and the turbulence-mixing model of the Todreas–Turi correlation model (TTM). Before simulating the thermal-hydraulics in FAIDUS, a few validations of DRM and TTM are conducted by comparing the numerical results of the pressure drop coefficients or temperature distribution obtained using ASFRE with the experimental data obtained using an apparatus with water or sodium for simulated FAs. After these validations, thermal-hydraulic analyses of FAIDUS and a typical FA are conducted for comparison. The numerical results indicate that, at a high flow rate simulating rated operation condition, no significant asymmetric temperature and velocity distribution occur in FAIDUS compared to the distribution in the typical FA. In addition, at a low flow rate simulating natural circulation condition in decay heat removal, the temperature distribution in FAIDUS is similar to that in the typical FA. This is because the local flow acceleration and the flow redistribution due to buoyancy force may occur in FAIDUS and the typical FA.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):021002-021002-15. doi:10.1115/1.4042356.

Laminar natural convection heat transfer from vertical 7 × 7 rod bundle in liquid sodium was numerically analyzed to optimize the thermal–hydraulic design for the bundle geometry with equilateral square array (ESA). The unsteady laminar three-dimensional basic equations for natural convection heat transfer caused by a step heat flux were numerically solved until the solution reaches a steady-state. The code of the parabolic hyperbolic or elliptic numerical integration code series (PHOENICS) was used for the calculation considering the temperature dependence of thermophysical properties concerned. The 7 × 7 heated rods for diameter (D =0.0076 m), length (L =0.2 m) and L/D (=26.32) were used in this work. The surface heat fluxes for each cylinder, which was uniformly heated along the length, were equally given for a modified Rayleigh number, (Raf,L)ij and (Raf,L)Nx×Ny,S/D, ranging from 3.08 × 104 to 4.28 × 107 (q =1 × 104∼7 × 106 W/m2) in liquid temperature (TL = 673.15 K). The values of ratio of the diagonal center-line distance between rods for bundle geometry to the rod diameter (S/D) for vertical 7 × 7 rod bundle were ranged from 1.8 to 6 on the bundle geometry with ESA. The spatial distribution of average Nusselt numbers for a vertical single cylinder of a rod bundle, (Nuav)ij, and average Nusselt numbers for a vertical rod bundle, (Nuav,B)Nx×Ny,S/D, were clarified. The average values of Nusselt number, (Nuav)ij and (Nuav,B)Nx×Ny,S/D, for the bundle geometry with various values of S/D were calculated to examine the effect of array size, bundle geometry, S/D, (Raf,L)ij and (Raf,L)Nx×Ny,S/D on heat transfer. The bundle geometry for the higher (Nuav,B)Nx×Ny,S/D value under the condition of S/D = constant was examined. The general correlations for natural convection heat transfer from a vertical Nx×Ny rod bundle with the ESA and equilateral triangle array (ETA), including the effects of array size, (Raf,L)Nx×Ny,S/D and S/D were derived. The correlations for vertical Nx×Ny rod bundles can describe the theoretical values of (Nuav,B)Nx×Ny,S/D for each bundle geometry in the wide analytical range of S/D (=1.8–6) and the modified Rayleigh number ((Raf,L)Nx×Ny,S/D = 3.08 × 104 to 4.28 × 107) within −9.49 to 10.6% differences.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):021003-021003-11. doi:10.1115/1.4041690.

When a depressurization accident of a very-high-temperature reactor (VHTR) occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, in order to predict or analyze the air ingress phenomena during a depressurization accident, it is important to develop a method for the prevention of air ingress during an accident. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Experiment and numerical analysis using a three-dimensional (3D) computational fluid dynamics code have been carried out to obtain the mixing process of two-component gases and the flow characteristics of localized natural convection. The numerical model consists of a storage tank and a reverse U-shaped vertical rectangular passage. One sidewall of the high-temperature side vertical passage is heated, and the other sidewall is cooled. The low-temperature vertical passage is cooled by ambient air. The storage tank is filled with heavy gas and the reverse U-shaped vertical passage is filled with a light gas. The result obtained from the 3D numerical analysis was in agreement with the experimental result quantitatively. The two component gases were mixed via molecular diffusion and natural convection. After some time elapsed, natural circulation occurred through the reverse U-shaped vertical passage. These flow characteristics are the same as those of phenomena generated in the passage between a permanent reflector and a pressure vessel wall of the VHTR.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(2):021004-021004-15. doi:10.1115/1.4041295.

Safety analyses at the high flux isotope reactor (HFIR) are required to qualify irradiation of production targets containing neptunium dioxide/aluminum cermet (NpO2/Al) pellets for the production of plutonium-238 (238Pu). High heat generation rates (HGRs) due to a fertile starting material (237Np), low melting temperatures, and previously unstudied material irradiation behavior (i.e., swelling/densification, fission gas release) require a sophisticated set of steady-state thermal simulations in order to ensure sufficient safety margins. Experience gained from previous models for preliminary target designs is incorporated into a more comprehensive production target model designed to qualify a target for three cycles of irradiation and illuminate potential in-reactor behavior of the target.

Commentary by Dr. Valentin Fuster

Technology Review

ASME J of Nuclear Rad Sci. 2019;5(2):024001-024001-8. doi:10.1115/1.4042194.
FREE TO VIEW

It is well known that electrical-power generation plays the key role in advances in industry, agriculture, technology, and standard of living. Also, strong power industry with diverse energy sources is very important for a country's independence. In general, electrical energy can be mainly generated from: (1) nonrenewable energy sources (75.5% of the total electricity generation) such as coal (38.3%), natural gas (23.1%), oil (3.7%), and nuclear (10.4%); and (2) renewable energy sources (24.5%) such as hydro, biomass, wind, geothermal, solar, and marine power. Today, the main sources for electrical-energy generation are: (1) thermal power (61.4%)—primarily using coal and secondarily using natural gas; (2) “large” hydro-electric plants (16.6%); and (3) nuclear power (10.4%). The balance of the energy sources (11.6%) is from using oil, biomass, wind, geothermal, and solar, and has visible impact just in a few countries. This paper presents the current status of electricity generation in the world, various sources of industrial electricity generation and role of nuclear power with a comparison of nuclear-energy systems to other energy systems. A comparison of the latest data on electricity generation with those several years old shows that world usage of coal, gas, nuclear, and oil has decreased by 1–2%, but usage of renewables has increased by 1% for hydro and 2% for other renewable sources. Unfortunately, within last years, electricity generation with nuclear power has decreased from 14% before the Fukushima Nuclear Power Plant (NPP) severe accident in March 2011 to about 10%. Therefore, it is important to evaluate current status of nuclear-power industry and to make projections on near (5–10 yr) and far away (10–25 yr and beyond) future trends.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2019;5(2):024501-024501-2. doi:10.1115/1.4042357.

Thallium bromide (TlBr) is a compound semiconductor material, which can be used for X-ray and gamma-ray detectors and can be used at room temperature. It has excellent physical properties, high atomic number and density, wide bandgap (B = 2.68 eV), and low ionization energy. Compared with other X-ray and gamma-ray detection materials, TlBr devices have high detection efficiency and excellent energy resolution performance. So TlBr is suitable for housing in small tubes or shells, and it can be widely used in nuclear material measurement, safeguards verification, national security, space high-energy physics research, and other fields. Based on the fabrication of TlBr prototype detector, this paper focuses on the device fabrication and signal acquisition technology. Gamma-ray spectrum measurements and performance tests are carried out with AM-241 radioactive source. The results show that the special photoelectric peak of 59.5 keV is clearly visible, and the optimal resolution is 4.15 keV (7%).

Commentary by Dr. Valentin Fuster

Discussion

ASME J of Nuclear Rad Sci. 2019;5(2):025501-025501-4. doi:10.1115/1.4042221.

In order to avoid the misuse of metal material in nuclear projects, typical cases happened in advanced passive pressurized water reactor (AP1000) nuclear power projects in China are analyzed. The analysis finding indicates that some cases were caused by defective procedures or undemanding processes performance, and all cases are found to be relevant to human error. It is considered that procedural management cannot completely avoid the misuse of metal material when it is caused by human error, and spectrometry analysis is suggested to reexamine the material of key components.

Commentary by Dr. Valentin Fuster

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