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Guest Editorial

Special Section on Research Center Řež: Nuclear-Engineering Activities in 2018

ASME J of Nuclear Rad Sci. 2019;5(3):030901-030901-6. doi:10.1115/1.4041277.

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030902-030902-8. doi:10.1115/1.4041564.

The austenitic steel 15-15Ti is being considered as one of the candidate materials for internal structural components of future heavy liquid metal (HLM) nuclear systems. This work studies the steel compatibility with liquid PbBi. Constant extension rate tensile (CERT) tests of tapered specimens were used to study sensitivity to liquid metal embrittlement (LME) and crack initiation. The taper creates a variation of stress along the gauge length which allows the identification of the stress and strain for the crack appearance. Testing was performed in air and in PbBi with 10−6 to 10−12 wt  % oxygen content at 300 °C. Post-test observation by scanning electron microscopy (SEM) highlighted the crack morphology. An evaluation of the environmental effect on the crack initiation is presented.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030903-030903-5. doi:10.1115/1.4041692.

Three-dimensional Monte Carlo code KENO-VI of SCALE-6.2.2 code system was applied for criticality calculation of the LR-0 reactor core. A central module placed in the center of the core was filled by graphite, lithium fluoride-beryllium fluoride (FLIBE), and lithium fluoride-sodium fluoride (FLINA) compounds. The multiplication factor was obtained for all cases using both ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. Obtained results were compared with benchmark calculations in the MCNP6 using ENDF/B-VII.0 library. The results of KENO-VI calculations are found to be in good agreement with results obtained by the MCNP6. The discrepancies are typically within tens of pcm excluding the case with the FLINA filling. Sensitivities and uncertainties of the reference case with no filling were determined by a continuos-energy version of the TSUNAMI sequence of SCALE-6.2.2. The obtained uncertainty in multiplication factor due to the uncertainties in nuclear data is about 650 pcm with ENDF/B-VII.1.

Topics: Uncertainty , Tsunamis
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030904-030904-4. doi:10.1115/1.4041790.

A granite block, acquired from a quarry Panská Dubenka located in the Czech Republic and used in presented experiments, is part of the same bedrock that can be potentially used for a deep geological repository. It is important to characterize advection in fractured rock to assess possible groundwater contamination. Newly used method—three-dimensional scanning using Hexagon Romer Arm was implemented to characterize the morphology of an examined fractured block with a aperture. The scanning technology provides the possibility to digitalize the rock surface. The scanning can be also used to determine any changes in the rock surface. The block was instrumented by tubing, and the aperture was sealed using a silicone. Flow paths were investigated by the comparison of fluid weights on the outlet on every output/site. The Hexagon Romer Arm is an ideal tool for the precise determination of a aperture's width in its full volume.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030905-030905-7. doi:10.1115/1.4042359.

The technology of molten salt oxidation (MSO) is a thermal treatment process mainly used for reprocessing of hazardous organic waste. This technology is considered as an alternative to the conventional incineration processes. The principle of the whole technology is based on flameless oxidation of materials in the molten salt with the consequent capture of gaseous products in molten alkaline salts. The melts with low melting point and high viscosity, such as a ternary mixture of carbonates Na2CO3, K2CO3, and Li2CO3, are the most used in this technology. However, the molten salts create a very corrosive environment for metal and ceramic materials, so the main aim of this experimental work was to determine the resistance of corundum samples, which were prepared by plasma spraying, and to find out its potential use as the protection of the reactor metal surface.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030906-030906-8. doi:10.1115/1.4042850.

This paper describes the measurement of 55Mn(n,2n) and 127I(n,2n) reaction rates in a well-defined reactor field in a special core of LR-0 reactor. The reaction rates were derived using gamma-spectrometry by measuring gamma activities of irradiated MnO2 and NaI samples at a high purity germanium (HPGe) detector. The spectral average cross section (SACS) in 235U prompt fission neutron spectrum (PFNS) was experimentally determined to be 0.2393 ± 0.015 × 10−3 b for 55Mn and 1.2087 ± 0.052 × 10−3 b for 127I. These obtained results were compared with calculations by MCNP6 code using ENDF/B VII.1, ENDF/B VII, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND- 2010, CENDL-3.1, and IRDFF nuclear data libraries. In a case of 55Mn, a good agreement with ENDF/B VII.1, JEFF 3.1, JENDL 3.3, JENDL 4, ROSFOND, and CENDL 3.1 nuclear data libraries was found, where C/E−1 is 0.1%, while IRDFF underestimated by about 15.8%. In the case of 127I, more significant discrepancies were found, where JENDL 3.3 and JENDL 4 overestimate the result by about 31.3%.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030907-030907-5. doi:10.1115/1.4043105.

This article presents experimental program, focused on extending lifetime and operation reliability of dissimilar welded joints of austenitic and perlite steel in conditions of secondary circuit media of water–water power reactor (VVER) unit. The main purpose of this experimental program is to bring new possibilities of corrosion protection of dissimilar welded joints, which are threatened by corrosion and degradation processes. These degradation processes significantly affect their service life. Suitability and corrosion behavior of thermal-sprayed coatings were tested.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030908-030908-5. doi:10.1115/1.4043103.

The article is primarily concerned with the measurement of heat recovery units for different work environments. At present, great emphasis is placed on indoor air quality. The quality of the indoor environment greatly influences the concentration, response, and overall quality of work. This is important in the area of the main control room of the nuclear power plant. For efficient suction of these spaces, it is necessary to handle the entire control process so that the air is delivered to the desired location in proper quantity and quality. For this reason, it is necessary to know the properties of the used components (e.g., recuperator). To calculate the efficiency of heat recovery unit, it is necessary to measure temperature, humidity, and air volume in its individual parts. Based on these requirements, a measurement system was created to achieve the appropriate data for the further optimization process. The first part briefly describes the created measuring system primarily tested on SL-Thermo (Plattling, Germany) units produced by the Südluft Systemtechnik (Plattling, Germany). The core of the system is based on an industrial programmable logic controller (PLC) in combination with LabVIEW (Austin, TX) software. The software can be used in laboratory condition or in the real operation. The next part is devoted to measurement of the heat recovery unit in laboratory condition, where the different operating states of the heat recovery unit were simulated. On the basis of the measurements made, a final evaluation was made and some structural modifications were recommended, which would increase the efficiency of the regenerative unit.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030909. doi:10.1115/1.4043099.

This paper presents results on the effect of a surface treatment on the environmentally assisted corrosion cracking in a pressurized water reactor chemistry. Slow strain rate testing of 316 L austenitic steel with selected rates was performed at pressurized water reactor (PWR) simulated water at 350 °C and in air at 300 °C. Detailed prior and post-testing characterization of two types of surfaces including roughness, hardness, and microstructural analysis was made. Transgranular cleavage-like environmentally assisted cracking (EAC) initiation and growth were observed under PWR conditions. The effect of two surface finishes on the cracking initiation was observed: (i) first crack initiates from the polished surface in a vicinity of the necking area rather than from the ground surface and (ii) then the deeper crack develops in the minimal diameter from the polished surface side than from the ground one.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030910-030910-7. doi:10.1115/1.4043097.

The molten salts can provide many possibilities for their use, e.g., an electrolyte in fuel cells or as a heat transfer medium and an oxygen transporter for flameless oxidation in molten salt oxidation (MSO) technology. The environment of molten salts is very corrosive; therefore, it is crucial to find such ceramic materials, which could be used as reactor filling for MSO technology. The aim of this work was to research physical properties of ceramic samples after the exposure within the eutectic mixture of Na2CO3, K2CO3, Li2CO3 and temperature of 700 °C.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030911-030911-7. doi:10.1115/1.4043199.

This article deals with issues arising during the design and production of a cold crucible (CC) for melting metals and alloys using electromagnetic induction. The article deals particularly with the results from tests and numerical simulations for designing the CC. The heat fluxes from different metals and their alloys to two different CCs and one calorimeter were measured during the tests. The required magnetohydrodynamic effects on the melted load were verified, and related (independent) electrical and thermal quantities were measured. The dependent electric parameters (R, L, Z) were measured on the inductor and on the primary side of the high frequency transformer. The experiments were numerically simulated first, and the experimental and simulated results were then compared. The final part of the article contains the final design of the CC. The final CC was tested for the transfer of energy from the inductor into a load placed inside the CC and the required magnetohydrodynamic effects on the melted load inside the CC were partly verified too.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030912. doi:10.1115/1.4043319.

We present a simple method of obtaining the turbulence kinetic energy spectrum from spatially resolved particle image velocimetry (PIV) data without need of use of Taylor hypothesis of frozen turbulence. Additionally, this method allows us to extract the velocity field components related to individual length scales resolved in the spectrum. We convolute the measured PIV velocity field with a band-pass filter, i.e., a difference of two Gauss functions with different widths, and the energy content of a such product we relate to the relevant length-scale of the used band. This is a “kind of wavelet transformation”, which, in respect to Fourier transformation, gives a good physical meaning to individual components.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030913-030913-9. doi:10.1115/1.4043378.

This work describes the computer model development of the water–water energetic reactor (VVER) 1000 nuclear power plant (NPP) in the methods for estimation of leakages and consequences of releases (MELCOR) 1.8.6 code and its subsequent use for the accident scenarios analysis leading to the core melting. The baseline accident scenario was a stress test case—the station blackout (SBO, the complete loss of alternating current electric power in a nuclear power plant). In addition to this, four other scenarios were analyzed in which the SBO was combined with other technological failures—the loss of steam generator feedwater system and small, medium, and large break coolant accidents (LOCA). The results provided detailed information on the time course of accident scenarios, their temperature and pressure parameters, hydrogen production, and the mass inventory released from the molten corium and debris into the containment of the NPP.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):030914-030914-8. doi:10.1115/1.4043377.

This paper focuses on the TRACE code assessment for helium-cooled systems thermal-hydraulic analysis. In the frame of the GoFastR (gas cooled fast reactor) European Collaborative Project, ENEA has offered some selected experimental data for the organization of a benchmark exercise aimed at the validation of the system and CFD codes for the gas reactor transient analyses. One of the Research Center Řež teams participated in it with a CFD code application. Now, the experimental data are used in order to assess the TRACE code for the ongoing high-temperature helium loop (HTHL-2) licensing process. The results of the TRACE calculations agreed very well with the experimental measurements (often within the experimental uncertainties) data provided by the He-FUS3 facility, indicating that the code, despite developed for water coolant applications, if an appropriately tuned input is adopted, it can also be suitable for reasonably accurate gas technology thermo-hydraulic simulations.

Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2019;5(3):031201-031201-11. doi:10.1115/1.4041432.

Corrosion behavior of 9%Cr F/M P92, E911, and EUROFER steels was investigated in flowing (2 m/s) Pb–Bi with 10−7 mass % O at 450 and 550 °C for up to 8766 and 2011 h, respectively. The steels show mixed corrosion modes simultaneously revealing protective scale formation, accelerated oxidation, and solution-based attack. At 450 °C, the accelerated oxidation resulted in a metal recession averaging 6 μm (± 2 μm) after ∼ 8766 h, while local solution-based corrosion attack ranged from ∼40 to 350 μm. At 550 °C, the accelerated oxidation resulted in a metal recession of about 10 μm (± 2 μm) after ∼ 2011 h. Solution-based corrosion attack appears more regularly at 550 °C, with a maximum depth ranged from ∼90 to 1000 μm. The incubation time for the solution based attack at 450 °C is 500–2000 h and < 300 h at 550 °C.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):031401-031401-9. doi:10.1115/1.4042021.

An analytical solution of the two-group diffusion equations is derived for multiregion source-driven subcritical systems in spherical geometry. An analytical formulation for the calculation of the effective multiplication factor is also presented. Using typical two-group cross sections characterizing source, buffer, and blanket regions, parameters such as neutron amplification, source efficiency, and blanket fast flux peaking factor are calculated. The criticality solution is utilized to calculate the effective multiplication factor and the neutron source efficiency. The dependency of the calculated global parameters on design variables like the source, buffer, blanket thicknesses, and subcriticality level is studied. Thin source regions result in very high neutron amplification, at the expense of high blanket fast flux peaking factors. If a buffer region is put between the source and the blanket regions, flux peaking could be reduced at the expense of reduced neutron amplification. If the subcriticality level can be reduced without jeopardizing safety, the neutron amplification increases and the fast flux peaking is reduced.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):031501-031501-8. doi:10.1115/1.4042115.

Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants (NPP). In the past few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus, structural integrity assessments considering PWSCC have become important. In this study, PASCAL-SP was improved considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate (CGR), and probability of crack detection. By using the improved version of PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were numerically evaluated. Moreover, the influence of leak detection and nondestructive examination (NDE) on failure probabilities was detected. Based on the obtained numerical results, it was concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):031601-031601-14. doi:10.1115/1.4042970.

This paper gives a complete account of recent work on the prediction of power restoration in severe accidents and events, and quantifies the benefits and impact of deploying back-up or emergency power systems in nuclear reactors. The overall outage data for all types of major events and disasters follow the same fundamental trends based on statistical learning theory, and are correlated by simple theoretically based exponential equations that include the degree of difficulty. This trend is shown to be completely independent of severe event type (e.g., hurricanes, ice storms, flooding, earthquakes, cyclones, and fires) but the rate systematically depends on the degree of difficulty. The paper emphasizes that the physics of learning, analytical methodology, technical statistical theory, and extensive event database already implicitly and fully include all human errors, actions, and decisions made during power restoration for the extremely adverse conditions prevalent in severe events and actual disasters. The theory shows flexible coping strategies/emergency power system FLEX/EPS reliability, deployment timescale, and severe event power restoration rate are intrinsically coupled together. The analytical results can be used to define the EPS design and reliability requirements and the potential risk benefit and deployment timescales in terms of the probability change in the risk of extended power loss.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):031701-031701-6. doi:10.1115/1.4042495.

The objective of the current work is to shed light on studying the air flow features of the air path which is part of the passive containment cooling system (PCS) in a pressurized water reactor design. A wind tunnel test using a 1:100 scaled model is established to study the characteristic called “wind-neutrality” of the air flow in the air path, which indicates that the environmental wind should not be beneficial or detrimental to the air flow for containment cooling. Test results show that the pressure distribution in the air path is uniform, and wind speeds, wind angles, and surroundings have little effect on air flow uniformity. These investigations show that it is possible to understand air flows in the air path of PCS with a scale wind tunnel test.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):031702-031702-10. doi:10.1115/1.4042541.

Control rod hydraulic drive system (CRHDS), which is invented by INET, Tsinghua University, is a new type of internal control rod drive technology. Control rod hydraulic deceleration device (CRHDD), which consists of the plug, the hydraulic deceleration cylinder, etc., is one of the main components of the CRHDS. The CRHDD performs the rod dropping deceleration function through the interworking of the plug and the deceleration cylinder, which is filled with water, and reduces the rod dropping peak acceleration and the impact force acting upon the control rod to prevent the control rod cruciform blade from being deformed or damaged. The working mechanism of the CRHDD is presented and analyzed. The theoretical model of the control rod dropping process, which is based upon force analysis of the control rod during scram process, three-dimensional flow field analysis, and flow resistance calculation of the hydraulic deceleration cylinder, the kinematics and dynamics analysis of the control rod, is built whose results are compared and validated by the CRHDS scram test results. Then the model is used to analyze the influence of the key parameters, including the fuel case gap, the plug design clearance, the working temperature, etc. on the CRHDD working performance. The research results can give guidance for the design and optimization of the CRHDD.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):031901-031901-4. doi:10.1115/1.4043096.

Boron carbide (B4C) particle-reinforced aluminum matrix composite is the key material for use as neutron absorber plate in fuel storage applications for Generation III advanced passive nuclear power plants in China. This material has once depended upon importing with various restrictions so that it has meaningful practical significance to realize the localized manufacturing for this material in China. More importantly, since it is the first time for this material to be used in domestic plant, particular care should be taken to assure the formal supplied products exhibit high stabilized and reliable service in domestic nuclear engineering. This paper initiates and proposes a principle design framework from technical view in qualification requirements for this material so as to guide the practical engineering application. Aiming at neutron absorber materials supplied under practical manufacturing condition in engineering delivery, the qualification requirements define B4C content, matrix chemistry, 10B isotope, bulk density, 10B areal density, mechanical property, and microstructure as key criteria for material performance. The uniformity assessment as to different locations of this material is also required from at least three lots of material. Only qualified material meeting all of the qualification requirements should proceed to be verified by lifetime testing such as irradiation, corrosion, and thermal aging testing. Systematic and comprehensive performance assessments and verification for process stabilization could be achieved through the above qualification. The long-term service for this neutron absorber material in reliable and safe way could be convincingly expected in spent fuel storage application in China.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2019;5(3):034501-034501-3. doi:10.1115/1.4039774.

This work deals with 23Na(n,2n)22Na and 127I(n,2n)126I reactions in the 252Cf spontaneous fission neutron source. 252Cf neutron source with approximate emission of 6·× 108 n/s was employed for the irradiation of sodium iodide. The spectrum-averaged cross sections (SACS) were then inferred from experimentally determined reaction rates and compared with calculations in MCNP6 using various nuclear data libraries. The experimental reaction rates were derived from the net peak areas (NPAs) measured using the high purity germanium spectroscopy. The measured SACS for the 23Na(n,2n)22Na reaction in the 252Cf spectrum was determined as equal to (8.98±0.32)·× 10−6 b. The resulting SACS for the 127I(n,2n)126I reaction in the 252Cf spectrum was derived as (2.044±0.0072)·× 10−3 b. These experimental data can be used for nuclear data libraries validation and to specify high energy tail of the 252Cf neutron spectrum.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):034502-034502-3. doi:10.1115/1.4042598.

This work is designed to artificially create test specimens with flaws that behave the same way as real-function flaws when observed by nondestructive testing (NDT) technologies. Thus, the understanding of the detection limitations of NDT methods is needed. In this study, real, realistic, and artificial flaws were compared by ultrasonic phased array technology. Fatigue flaws, which belong to the most common structural issues (Ruzicka, M., Hanke, M., and Rost, M., 1987, Dynamicka Pevnost a Zivotnost, CVUT, Prague, Czech Republic, p. 75), are investigated. Measurements have revealed significant differences in the amplitude of ultrasonic echo from fatigue cracks in distinct phases of crack propagation. Studied specimens with realistic flaws have demonstrated their quality for calibration, staff training, and NDT system qualification. More realistic test specimens will increase ultrasonic test result reliability.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):034503-034503-5. doi:10.1115/1.4042704.

Algorithm library plays an important role in digital instrument control system. In the design process, software testing and safety verification are focused on safety and reliability of the algorithm library. The configured parameters are important for the safety and reliability. Relations of parameters among different modules are very complex. Parameters are easy to be wrongly configured in the design process. Parameters must be considered and checked. The analysis process of safety constraints is established. The analysis is not only important for the sufficient in the design validation and verification but also improves design quality and decreasing the defects.

Topics: Safety , Signals , Algorithms
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2019;5(3):034504-034504-6. doi:10.1115/1.4043462.

Nuclear grade fuel claddings act as a barrier against release of fuel particles into the coolant water during nuclear power plant operation, handling, and wet or dry storage of the spent fuel rods. The integrity of claddings is always a critical issue during the reactor operation, storage, or loss of coolant accidents (LOCA). After Fukushima disaster, cladding materials are widely studied, as the neutron transparent zirconium (Zr)-based alloys or accident tolerant fuel claddings, which should reduce the high-temperature oxidation rate of Zr-based alloys and enhance the accident tolerance. One part of the long-term investigation of Zr-based claddings in the Research Center Řež contributes to the characterization of material behavior after creep testing and high-temperature oxidation, i.e., conditions that may occur during reactor operation or storage of the spent fuel. Then, the microstructure response can be investigated by means of standard as well as advanced microscopy methods. In this study, we introduce the methodology of the cladding microstructure complex characterization demonstrated on the reference (nonirradiated) Zr-1%Nb alloy after creep testing. In the Zr-1%Nb microstructure, the hydrogen is present under some conditions as the hydride phases ZrH or ZrH2 that cause the mechanical properties degradation. As well, the hydride reorientation can pose a risk. Therefore, currently used methods of hydrides evaluation are presented. Furthermore, we introduce the microstructure analysis of the zircalloy-4 cladding after high-temperature oxidation at the temperature 1300 °C simulating LOCA. For this study, the new automatic chemical analysis procedure was developed. The main issue is the determination of the oxygen concentration and alloying elements behavior after critical increase of oxygen in the Zr-alloy tube at LOCA, also concerning the hydrogen content in the microstructure. Based on the element concentration gradients, the phase transitions from the oxide ZrO2 to the prior β-Zr can be determined and contributes to the phase diagrams experimental verification. In our contribution, the significant role of the electron microscopy methods “from micro to nano-scale” in the nuclear research is emphasized in these two research topics.

Commentary by Dr. Valentin Fuster

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