Reactor Irradiation Research Articles

Experimental study on influence of shielding gas pressure on the quality of pulsed laser welding of thin walled T91 steel clad tube with end plug

Fabrication of metal fuel pins for test irradiation in fast breeder test reactor requires welding of thin wall clad tube with solid end plug of modified 9Cr-1Mo steel material under high pressure helium atmosphere. Autogenous pulsed Gas Tungsten Arc Welding [GTAW] technique is currently used for end cap welding of metal fuel pins inside the glove box. Pulsed laser welding of end plug with clad tube of modified 9Cr-1Mo steel material under inert atmosphere at 1.2 bar gas pressure is under development.Pulsed laser beam welding [PLBW] is considered as alternate welding technique to replace the pulsed tungsten inert gas welding.In the current study the influence of shielding gas pressure on the weld joint characteristics like depth of penetration, bead width, and heat affected zone were investigated to optimize laser welding parameters. Initially experiments were conducted under helium gas at atmosphere pressure to optimize the process parameters to obtain the desired depth of penetration and then repeated the experiment with the optimized parameters at 0.5 bar pressure to study the effect of shielding gas pressure. Laser beam welding parameters like peak power − 1800 to 2400 W, pulse frequency – 5 to 8 Hz and pulse duration – 6 to 10 ms were varied during welding sample coupons. Welded sample coupons were subjected to various NDT techniques and subsequently metallographic examination was carried out to measure the bead width, depth of penetration and delta ferrite distribution in the weld metal. Laser beam welding parameters were optimized at atmosphere pressure to meet the stringent acceptance criteria. Further study was carried out under helium gas atmosphere at 0.5 bar using optimized welding parameters and it reveals that the drastic reduction in heat input due to laser beam attenuation and absorption by quartz window.

Annealing influence on the microstructure of irradiated U-Mo monolithic fuel foils

In this study we compared the microstructure evolution of U-Mo fuel foils produced with and without heat treatment at low burn-up via advanced post-irradiation examination. The aim of this study is to observe after irradiation the ways in which the fabrication processes have influenced fuel behavior at early-stage irradiation, as for very low burn up microstructural studies are lacking. In this work it was observed that the larger grain size detected in the heat-treated samples before irradiation led to decreased grain refinement after irradiation. Grain refinement was associated with the presence of small nano-size bubbles and precipitates. This phenomenon is hypothesized to influence early fuel swelling during reactor irradiation. Grain refinement was also observed to increase in regions where γ-U decomposition was present. Thus, it was enhanced in the samples fabricated without heat treatment. The heat treatment also increased the thickness of the U-Mo/Zr interface, namely of the UZr2 layer. However, the influence of this layer on fuel performance needs further investigation. On one side, it may contribute to better mechanical bonding; on the other, it may influence swelling and blistering in the interaction layer as porosity increases when this layer is increased. This was observed especially in the presence of increased area containing low Mo concentration, and thus containing a higher fraction of the α-U phase, which is highly susceptible to irradiation induced swelling. Strong evidence of reverse transformation under irradiation (α-U + γ′-U2Mo → bcc γ-U) was observed in these samples. While the precipitates (carbides and oxide) seem to be unaffected by the irradiation at these low burnups. However, further analyses are necessary at higher burn-up to assess the exact impact different heat treatments have on fuel performance.

An in-pile demonstrated platform for the evaluation of tritium extraction technologies in China Mianyang Research Reactor

The preliminary design of the tritium extraction system for CFETR breeding blanket has been completed. In the design, two molecular sieve packed columns are connected in series to achieve both hydrogen-helium separation and tritium enrichment. The first column primarily focuses on hydrogen-helium separation and volume reduction, while the second column, which is thinner and longer than the first, is mainly used for tritium enrichment. To achieve tritium-helium separation and tritium storage simultaneously, a U-shaped tritium storage bed is utilized. To verify the feasibility of these tritium extraction technologies, an in-pile tritium release and recovery platform was constructed in China Mianyang Research Reactor (CMRR). This paper provides a detailed description of the in-reactor irradiation device and the design and composition of tritium extraction system. Moreover, an in-situ experiment was conducted in early 2022 to produce and recover tritium with a total activity level of approximately one hundred curies. The experiment obtained various parameters, including tritium release form and ratio of Li4SiO4 with high 6Li abundance, tritium extraction rate, tritium recovery rate of the whole system and tritium penetration in 316 L stainless steel. These results allow for a preliminary evaluation of the feasibility of the designed tritium extraction technologies.

Application of ion implantation as a tool to study neutron induced morphological changes in HOPG and RBMK-1500 reactor graphite

Based on the RBMK-1500 (rus.: peaктop бoльшoй мoщнocти кaнaльный, PБMК; reaktor bolshoy moshchnosti kanalnyy, “high-power channel-type reactor”) reactor irradiation conditions, ion implantation was used as a tool to study neutron induced morphological changes in both highly oriented pyrolytic graphite (HOPG) and nuclear grade RBMK graphite. Graphite samples were implanted with 180 keV 14N+ ions at the fluences of 1.0 × 1016 ions/cm−2 and 2.5 × 1016 ions/cm−2. To study temperature effects on both ion migration and structural disorder level in the graphite matrix, the implantation procedures were carried out under different temperature conditions – room temperature (RT) and 500 °C. Subsequently, the distribution profiles of the implanted 14N+ ions were obtained by using secondary ion mass spectroscopy (SIMS) technique, while the microstructural properties of the graphite were evaluated by Raman spectroscopy. Implantation induced primary displacement damage profiles and implanted ion profiles were evaluated theoretically by using GEANT4 10.6 and SRIM-2013 codes. The profiles of implanted nitrogen obtained by SIMS technique were found to be in good agreement with theoretical ones. The surfaces of HOPG samples displayed macroscopic damage in form of fractures after 14N+ ion implantation at 500 °C; however, in case of RBMK graphite, the sample surfaces remained without visually observable changes. The Raman spectra showed an increase of sp3-related content and formation of amorphous carbon in both HOPG and RBMK graphite samples; the latter exhibits higher structural disorder at microscopic level and stronger amorphization than HOPG.

Modified Ultrasonic Irradiation reactor: Application on Produced Water Treatment

Modified Ultrasonic Irradiation reactor: Application on Produced Water Treatment

Production of 192Ir sealed sources in Es Salam research reactor for non-destructive testing

Abstract An iridium-192 miniature sealed source was produced by irradiation of 12 metal disks of natural iridium (470.5 mg) under a thermal neutron flux of 9.4 × 1013 n cm−2 s−1 during 200 h in Es Salam research reactor (Algeria). The preliminary activity was measured after 10 days decay time by a CAPINTEC type CRC-712 M dose calibrator, the given activity was 1.9 × 1011 Bq (5.2 Ci). To estimate source activity, a modeling of reactor core and irradiation container with the targets was made by MCNP5 code with and without iridium targets to evaluate the effect of the disturbance and reaction; the calculated activity using disturbed flux was comparable with that measured. A non-destructive test using 192Ir sealed sources with GammaMat TSI type B(U) projectors was carried out to evaluate the integrity of the weld joint at the bend of a carbon steel pipe. The weld quality of two plates; carbon steel and stainless steel was also evaluated. In terms of sensitivity, the resulting gamma images matched well and reported a real finding on the state of tested samples.

Radiation shielding optimization design research based on bare-bones particle swarm optimization algorithm

In order to further meet the requirements of weight, volume, and dose minimization for new nuclear energy devices, the bare-bones multi-objective particle swarm optimization algorithm is used to automatically and iteratively optimize the design parameters of radiation shielding system material, thickness, and structure. The radiation shielding optimization program based on the bare-bones particle swarm optimization algorithm is developed and coupled into the reactor radiation shielding multi-objective intelligent optimization platform, and the code is verified by using the Savannah benchmark model. The material type and thickness of Savannah model were optimized by using the BBMOPSO algorithm to call the dose calculation code, the integrated optimized data showed that the weight decreased by 78.77%, the volume decreased by 23.10% and the dose rate decreased by 72.41% compared with the initial solution. The results show that the method can get the best radiation shielding solution that meets a lot of different goals. This shows that the method is both effective and feasible, and it makes up for the lack of manual optimization.

Radiation-induced structure of austenitic steels with different nickel content under neutron irradiation in SM-3 and BOR-60 reactors

Comparative studies of the radiation-induced structure of austenitic steels with a nickel content of 10, 20 and 25 wt.%, irradiated sequentially in the SM-3 and BOR-60 reactors, as well as to higher damaging doses in the BOR-60 reactor, have been carried out. The phase composition, dislocation structure, pores, and radiation-induced segregations at grain boundaries were studied by high-resolution analytical methods of transmission electron microscopy, scanning electron microscopy, and atomic probe tomography. The formation of radiation-induced phase precipitates based on nickel has been established, and its volume fraction correlates with the level of radiation-induced segregations, and increases, the higher the nickel content in the steel. The values of barrier strength factors for radiation-induced structural elements in the studied steels are adjusted by calculation and experiment, which makes it possible to determine their contribution to radiation hardening.It is shown that the largest contribution to radiation hardening as a result of neutron irradiation in BOR-60 at high irradiation temperature up to 29 dpa is made by large radiation-induced precipitates of (G + γ') phases. It is shown that with an increase in the damaging dose, the main factor limiting the performance of internal devices will be radiation swelling, since the contribution to the change in properties from radiation-induced phases and radiation defects will not increase due to their density reaching saturation. Steel with 25 wt.% Ni exhibits the lowest level of swelling at high radiation doses, which makes it possible to consider it as a material-candidate for internals for promising VVER reactors with higher temperatures and longer service life.

X-ray diffraction line broadening and radiation damage

The diffraction line profiles of single crystal Zr and Zr-2.5Nb pressure tubing have been measured and analysed before and after irradiation in a nuclear reactor. The irradiation produces damage in the form of prismatic dislocation loops. In all cases, Fourier analysis of the deconvoluted peak using the method of Warren and Averbach shows that the magnitude of the distortion (micro-strain) component is not consistent with the dislocation density measured by transmission electron microscopy. The discrepancy can be related to the reduced strain energy from the dipoles on opposite sides of the prismatic dislocation loops and loop alignment, which reduces the average strain in the same way as polygonization or sub-grain boundary formation. The Fourier analysis also gives a value for the coherent diffraction domain size that is consistent with the mean spacing between the dislocations and thus the dislocation density.

Influence of plutonium content in dissolver solutions derived from irradiated fast reactor fuels on plutonium stripping in multistage countercurrent liquid–liquid extraction with acid split flowsheet

ABSTRACT In the acid split flowsheet, the Pu content in the dissolver solution of nuclear fuel has a significant impact on the configuration of flowsheet and the Pu content in the products. The influence of Pu content in dissolver solutions of nuclear fuel on the Pu stripping has been studied. The decontamination performance of Cs as major fission products was assessed, and Cs was decontaminated well in the acid split flowsheet. High HNO3 concentration in the feed solution and at the co-decontamination section was adjusted for the co-recovery of Np with U and Pu, contributing to Np oxidation and the increase in the distribution ratio of Np. The experimental results demonstrate that the Pu content in the U/Pu and U products increased with an increase in the Pu content in the feed solution. The calculated results indicate that the Pu leakage into the U product is suppressed with the Pu stripping solution only at low temperature under our experimental conditions, and Pu is recovered into the U/Pu product efficiently. This study proves that the Pu leakage into the U product is controlled and the acid split flowsheet is flexible with change in the Pu content in the feed solution.

Thermal-Hydraulic Analysis of the AFIP-7 Irradiation Test in the Advanced Test Reactor—Oxide Growth Prediction and Correlation

Knowing the thickness of the oxide layer on the surface of aluminum fuel cladding is vitally important for predicting fuel temperature due to the low thermal conductivity of the oxide layer. Several correlation models for predicting oxide growth can be found in the literature. In previous research, the correlations were combined with heat transfer simulations in Abaqus, a finite element analysis code, to forecast the oxide growth. However, this approach requires heat transfer coefficients for modeling heat exchanges with the external flow field, and such coefficients were obtained through empirical equations. Since different empirical equations yield varying heat transfer coefficients, the cladding temperature and predicted oxide thickness both carry a high degree of uncertainty. This research develops a new approach that integrates the fluid flow, fluid and solid heat transfer, and oxide growth correlation(s) into a single computational fluid dynamics model. We demonstrate this approach’s ability to predict oxide development on the AFIP-7 plates during two Advanced Test Reactor (ATR) irradiation cycles. The projected oxide thickness falls within the experimental measurements taken during post irradiation examination.

Understanding the effect of specimen size on the properties of fine-grain isotropic nuclear graphite for irradiation studies: Mechanical properties

Programs at Oak Ridge National Laboratory are investigating the effects of neutron irradiation in the High Flux Isotope Reactor (HFIR) on properties of fine-grain isotropic nuclear graphites. Specimens were irradiated in the HFIR flux trap, which requires using specimens that are smaller than the sizes recommended by ASTM standards. This work has investigated whether subsized specimens can be used to reliably predict strength behavior. The results show that (1) tensile strength was loosely coupled to specimen volume, (2) the compression specimens had a slightly higher strength for larger volumes, (3) the uniaxial flexural strength was nearly independent of specimen effective area for one orientation whereas the other orientation showed a slight strength–size scaling Weibull behavior, and (4) equibiaxial flexural strength followed traditional strength–size scaling Weibull behavior. These results suggest that fine-grain graphite is more similar to ceramics than to medium- and large-grained graphite with respect to both the strength-limiting flaws (Griffith theory) and the effect of machining on the strength behavior. This work shows that using specimens that are smaller than ASTM recommendations are suitable for investigating the strength changes caused by neutron irradiation because the primary interest is the relative change in the strength distribution parameters rather than their absolute values.

Photocatalytic Degradation of Crystal Violet Dye under Visible Light by Fe-Doped TiO2 Prepared by Reverse-Micelle Sol-Gel Method.

A reverse-micelle sol-gel method was chosen for the preparation of Fe-doped TiO2 samples that were employed in the photodegradation of the crystal violet dye under visible light irradiation in a batch reactor. The dopant amount was varied to assess the optimal photocatalyst composition towards the target dye degradation. The photocatalysts were characterized through a multi-technique approach, envisaging XRPD and QPA as obtained by Rietveld refinement, FE-SEM analysis, DR UV-vis spectroscopy, N2 adsorption/desorption isotherms measurement at -196 °C, ζ-potential measurement, and XPS analysis. The physical-chemical characterization showed that the adopted synthesis method allows obtaining NPs with uniform shape and size and promotes the introduction of Fe into the titania matrix, finally affecting the relative amounts of the three occurring polymorphs of TiO2 (anatase, rutile and brookite). By increasing the Fe content, the band gap energy decreases from 3.13 eV (with undoped TiO2) to 2.65 eV (with both 2.5 and 3.5 wt.% nominal Fe contents). At higher Fe content, surface Fe oxo-hydroxide species occur, as shown by DR UV-vis and XP spectroscopies. All the Fe-doped TiO2 photocatalysts were active in the degradation and mineralization of the target dye, showing a TOC removal higher than the undoped sample. The photoactivity under visible light was ascribed both to the band-gap reduction (as confirmed by phenol photodegradation) and to dye sensitization of the photocatalyst surface (as confirmed by photocatalytic tests carried out using different visible-emission spectra LEDs). The main reactive species involved in the dye degradation were determined to be positive holes.

Cold brittleness of metals as a structural multistage dislocation process

The cold brittleness of metals under conditions "before --- during --- after" irradiation in nuclear and thermonuclear reactors is considered. The conditions and mechanisms of cold brittleness are determined depending on the type of their crystal lattices, structures, physical and mechanical properties and external influences (temperature, stress, reactor irradiation). The mechanisms of cold brittleness are dislocational and define cold brittleness as a structural multistage process. Stress relaxation at the front of a subcritical crack stops its growth (cold brittleness is not formed). The stacking fault energy of the metal determines the critical level of stress for the beginning of its relaxation by mechanical twinning at the crack front (external stress concentrator). In irradiated metals (cold-brittle "before" irradiation), the temperature range of cold-brittleness expands. "During" irradiation, cold brittleness is not formed. Keywords: alloys, steels, low temperatures, reactor irradiation, crack nucleation, brittle fracture, stages, mechanisms.

Хладноломкость металлов как структурный многостадийный дислокационный процесс

COLD BRITTLENESS OF METALS AS A STRUCTURAL MULTISTAGE DISLOCATION PROCESS V.M. Chernov1,2 1A.A. Bochvar High-Technology Research Institute of Inorganic Materials, Moscow, Russia. 2National Research Nuclear University MEPhI, Moscow, Russia. E-mail: VMChernov@bochvar.ru. The cold brittleness of metals under conditions "before - during - after" irradiation in nuclear and thermonuclear reactors is considered. The conditions and mechanisms of cold brittleness are determined depending on the type of their crystal lattices, structures, physical and mechanical properties and external influences (temperature, stress, reactor irradiation). The mechanisms of cold brittleness are dislocational and define cold brittleness as a structural multistage process. Stress relaxation at the front of a subcritical crack stops its growth (cold brittleness is not formed). The stacking fault energy of the metal determines the critical level of stress for the beginning of its relaxation by mechanical twinning at the crack front (external stress concentrator). In irradiated metals (cold-brittle “before” irradiation), the temperature range of cold-brittleness expands. "During" irradiation, cold brittleness is not formed.

Multi-group analysis of Minor Actinides transmutation in a Fusion Hybrid Reactor

New nuclear technologies are currently being study to face High Level Waste treatment and disposal issues. Generally, GEN-IV fission Fast Reactors (FR) are considered the waste-burners of the future. In fact, a fast flux turns out to be the best choice for actinides irradiation in critical reactors because of favorable cross section conditions. Differently, Fusion Fission Hybrid Reactors (FFHR) are futuristic devices based on the combination of fusion and fission systems and could represent an alternative to FRs. In such systems, the choice spectrum of the neutron flux that irradiates HLW may be non-obvious due to some operational constraints which have to be considered. To design and optimize these systems as waste-burners, one should fully understand the transmutation dynamics occurring into the fission region. A multi-energy-group analysis by FISPACT-II code has been set to analyze the conversion processes in scenarios characterized by different neutron energy spectra and fluences. The results of this study show that, despite fast fluxes are characterized by better behaviors in terms of radiotoxicity treatment, the difficulties of reaching high reaction yields may require solutions involving moderators or broadened neutron fluxes to increase the reactions probabilities and, consequently, actinides mass conversion yield.

Heat Transfer Characteristics for Multi-Silicon Ingots Irradiation in a Typical Research Reactor

Heat Transfer Characteristics for Multi-Silicon Ingots Irradiation in a Typical Research Reactor

Simulation of lutetium metal evaporation with considering atomic metastable state

Lutetium-177, one of the current global research hotspots for medical isotopes, can be produced by reactor irradiation of lutetium-176, which can be produced and processed with the assistance of the electron beam physical vapor deposition (EB-PVD). The metal evaporation process is an important part of EB-PVD, which aims to convert solid metal into metal atomic vapor. The distribution of each macroscopic characteristic quantity, such as thermal population distribution, density distribution, velocity distribution, and temperature distribution of metal atomic vapor, will directly affect the utilization of metal vapor atoms, the subsequent deposition process, and coating process. In this paper, two-dimensional model and three-dimensional model of the metal evaporation process are established based on the direct simulation Monte Carlo method, and atomic metastable state is introduced into the collision assumption. The macroscopic characteristic quantities, such as the number density of metal atomic vapor, the motion velocity and the discrete degree of the total motion velocity in the evaporation process with or without considering the lutetium atomic metastable state, are analyzed and compared. The spatial distribution characteristics of each energy level of metal atomic vapor are observed, and the variation characteristics of the macroscopic characteristic quantities in the center with the slit opening size and the evaporation source surface temperature are discussed. The study results show that the number density of metal vapor passing through the beam flow device decreases, the average velocity in the <i>x</i>-direction and <i>z</i>-direction and the discrete degree of the total motion velocity increase, with the atomic metastable state taken into consideration. The beam flow device intensifies the transformation of the atomic energy level of the metal vapor and changes the physical parameters of the metal atomic vapor. At the same time, the slit opening size and the evaporation source surface temperature trigger off the change of the macroscopic characteristic quantity distribution of the metal vapor, including the degree of anomalous protrusion of each macroscopic characteristic quantity above the slit. The results of the three-dimensional simulation based on the two-dimensional model are consistent with the changes in the order of magnitude of data distribution and spatial distribution pattern, which can serve as a reference and guidance for the experimental study of the electron beam metal evaporation process.

Biorreactores empleados para la biorremediación de aflatoxinas

The toxicity of microorganisms such as aflatoxins found in various varieties of food that are currently widely consumed in Peru as in the world. Many of the foods contaminated with aflatoxins cause diseases, one of the main ones being carcinogenicity. This study aimed to evaluate the effect of ozone/air and UV concentration on the remediation of aflatoxins in different cereals or grains for human consumption. This study reviews how a new reactor can be developed that degrade and detoxify into more aflatoxins, however, many studies have verified that the harmful effects of AFB1 can be greatly reduced by UV irradiation in the photodegradation reactor, and the reactor can be applied on a large scale in the detoxification of AFB1 of many kinds of cereal, as well as in sewage. Therefore, there has been a growth in the development of new biofilm-type and closed reactors to reduce the amount of this toxin in food. The creation of new bioreactors such as the fluidized bed one is very promising for the decontamination of aflatoxins in cereals and would be less dangerous for consumers.

Light-enhanced thermochemical production of solar fuels from methane via nickel-based redox cycle

Thermochemical methane reforming driven by solar energy is one of the efficient ways to store the renewable energy and to simultaneously reduce greenhouse gas emission. We report reaction reactivity of steam reforming of methane using nickel-based oxygen carrier driven by pure thermal heating and direct light irradiation in the same reactor. We find that the light-driven steam methane reforming system show improved conversion and reaction rate compared to that driven by pure thermal heating. We observed that illuminated NiO/NiAl2O4 achieved the same methane conversion rates and the H2 production rate comparable to those obtained on the same samples operated with only a thermal energy input, but at temperatures up to ∼ 50 K lower. Based on in situ diffuse reflectance Infrared Fourier transform spectroscopy (DRIFTs) experiments, we postulate that light on the nickel oxide surface act to activate CH4 desorption and facilitate CHxO species evolution. By the combined effects of selection of chemical looping reactions and operation conditions, the solar-to-fuel efficiency of 16% was achieved. The results showcase the design of more energy-efficient chemical processes with a significant fraction of energy input provided in the form of direct sunlight irradiation as a potent pathway that can complement extremely high-temperature thermochemistry efforts in the quest for more efficient solar fuel production.