Reactor Irradiation Research Articles

Sub-size tensile specimen design for in-reactor irradiation and post-irradiation testing

The present work aims to provide a complete engineering solution, an appropriate experimental database, and a brief physical background on designing a miniature specimen geometry suitable for irradiation in materials test reactors such as the High Flux Isotope Reactor and post-irradiation out-of-hot cell testing. The physical limits of specimen miniaturization and a background of the scale factor effect are discussed, and principal limitations are defined. The advantages of modern test methods like digital image correlation, as well as some limitations connected to small specimen size, are analyzed. A sub-sized specimen geometry, “SS-Mini,” is designed; the geometry employs existing irradiation capsules leading to the reduced cost of any irradiation campaign. The new geometry performance is evaluated using a commercial 304L stainless steel, an aluminum alloy including advanced 3D-printed material, a high nickel 718-alloy, tungsten, and an advanced fuel cladding FeCrAl alloy. Mechanical tests are conducted to compare the engineering mechanical properties (yield and ultimate tensile stress, uniform and total elongation values) and plastic behavior of the proposed miniature specimen with common specimen types for irradiation testing.

Carbon vacancy-induced enhancement of the visible light-driven photocatalytic oxidation of NO over g-C3N4 nanosheets

g-C3N4 (gCN) with carbon vacancy has been extensively investigated and applied in (photo)catalysis. Engineering the carbon vacancy in gCN is of great importance, but it remains a challenging task. In this work, we report for the first time the fabrication of gCN with carbon vacancy (Cv-gCN) via thermal treatment of pristine gCN in CO2 atmosphere. The photocatalytic performance of Cv-gCN is evaluated on the basis of NO oxidization under visible light irradiation (λ>400nm) in a continual reactor. The successful formation of carbon vacancy in gCN is confirmed through electron paramagnetic resonance (EPR) and X-ray photoelectron spectroscopy (XPS). The photocatalytic oxidation removal rate of NO over Cv-gCN is 59.0%, which is two times higher than that over pristine gCN (24.2%). The results of the quenching experiment show that superoxide radicals (O2−) act as the main reactive oxygen species, which is responsible for the oxidation of NO. The enlarged BET surface areas and negatively shifted conduction band (CB) potential enhance the photocatalytic activity of Cv-gCN, which facilitates the efficient electron transfer from the CB of Cv-gCN to the surface adsorbed oxygen, resulting in the formation of O2− that can oxidize NO.

Computer modeling of thermal processes involving calcium, strontium and cesium during radioactive graphite heating in the carbon dioxide atmosphere

The design of nuclear reactors did not provide for decommissioning solutions, and there were no safe technologies to handle irradiated reactor graphite. Decommissioning of uranium–graphite reactors is a combination of complex tasks involving the selection of appropriate methods and techniques for the radioactive graphite handling. Computer modeling of the reactor graphite reprocessing by heating in a carbon dioxide environment makes it possible to determine the behavior of radioactive elements. The behavior of Ca, Sr and Cs during radioactive graphite heating in the carbon dioxide atmosphere was studied through computer modeling. It has been found that calcium is present as vapors of Ca, CaO, CaCl and CaCl2, as Ca+ and CaO+ ions, and as condensed forms of CaCO3, CaCl2 and CaO. Strontium is present as vapors of Sr, SrO, SrCl and SrCl2, as Sr+ and SrO+ ions, and as condensed forms of SrCl2, SrCO3 and SrO. Cesium is present as vapors of Cs and CsCl, as Cs+ ions, and as a condensed phase of CsCl2. Basic reactions have been identified and their respective equilibrium constants have been determined. The data obtained has shown that formation of calcium, strontium and cesium chloride vapors takes place at temperatures of 573 to 973K. A temperature increase to 1373K leads to thermal ionization of cesium chloride and to formation of ionized cesium. As the temperature increases to 2273K, thermal ionization of strontium and cesium is observed and ionized calcium and strontium form.

Structural and thermodynamic study of dicesium molybdate Cs2Mo2O7: Implications for fast neutron reactors

The structure of α-Cs2Mo2O7 (monoclinic in space group P21/c), which can form during irradiation in fast breeder reactors in the space between nuclear fuel and cladding, has been refined in this work at room temperature from neutron diffraction data. Furthermore, the compounds' thermal expansion and polymorphism have been investigated using high temperature X-ray diffraction combined with high temperature Raman spectroscopy. A phase transition has been observed at Ttr(α→β)=(621.9±0.8)K using Differential Scanning Calorimetry, and the structure of the β-Cs2Mo2O7 phase, orthorhombic in space group Pbcm, has been solved ab initio from the high temperature X-ray diffraction data. Furthermore, the low temperature heat capacity of α-Cs2Mo2O7 has been measured in the temperature range T=(1.9–313.2)K using a Quantum Design PPMS (Physical Property Measurement System) calorimeter. The heat capacity and entropy values at T=298.15K have been derived as Cp,mo(Cs2Mo2O7,cr,298.15K)=(211.9±2.1)JK−1mol−1 and Smo(Cs2Mo2O7,cr,298.15K)=(317.4±4.3)JK−1mol−1. When combined with the enthalpy of formation reported in the literature, these data yield standard entropy and Gibbs energy of formation as ΔfSmo(Cs2Mo2O7,cr,298.15K)=−(628.2±4.4)JK−1mol−1 and ΔfGmo(Cs2Mo2O7,cr,298.15K)=−(2115.1±2.5)kJmol−1. Finally, the cesium partial pressure expected in the gap between fuel and cladding following the disproportionation reaction 2Cs2MoO4=Cs2Mo2O7+2Cs(g)+ 1/2 O2(g) has been calculated from the newly determined thermodynamic functions.

Repeatability of irradiation damage and of recovery by post-irradiation annealing of EUROFER base steels

Former investigations clearly revealed that embrittlement and hardening of reduced-activation ferritic-martensitic (RAFM) steels after neutron irradiation can be reduced remarkably by post-irradiation annealing (PIA). This study demonstrates the repeatability of damage and recovery mechanisms for EUROFER-based and EUROFER-ODS-HIP-steels by using specimens of previous 330°C Bor 60 reactor (SSC RIAR) irradiations. Specimens were investigated in the irradiated state after a first PIA, after a subsequent further irradiation, and after a second annealing.Both tensile and Charpy-V impact tests clearly indicate that embrittlement occurs after the second irradiation of intermediately annealed (and, thus, recovered) material in the same order as after the first irradiation.In addition, a subsequent PIA leads to a complete recovery of the newly applied neutron damage, suggesting that lifetime of neutron-exposed fusion power plant components could be extended substantially by means of intermediate annealing.

Biokinetic study of selenium nanoparticles and salt forms in living organisms

An experimental study of absorption, distribution, metabolism, and excretion from the body of radioisotope-labeled nanoparticles (NPs) of elemental Se in comparison with the traditional form of this trace element (sodium selenite) has been performed, being administered in the gastrointestinal tract of rats that were normally supplied with Se or experiencing its nutritional deficiency. Nuclear activation analysis is used for the detection of Se in the biological tissue, which is based on the measurement of a gamma-emitting label [75Se34] that is introduced in the Se NPs or its salt by thermal neutron irradiation in a nuclear reactor. It is shown that Se NPs administered into the gastrointestinal tract have bioavailability comparable to the salt form of this element. According to some experimental data, the metabolic assimilation of both forms of Se is greater in animals that suffer from a deficiency in this element when compared to those that are normally supplied with it. There are some differences in the kinetics of accumulation of Se NPs and its salts in the blood and liver, which can be explained by the presence of limiting steps in the absorption and biotransformation of Se-containing NPs. The retention of Se that is administered in the form of NPs or salts to Se-deficient animals differs significantly. These results demonstrate the capacity of the Se NPs to be a source of Se in a new generation of biologically active food supplements.

Investigation of hydrogen isotopes interaction with lithium CPS under reactor irradiation

This work describes the experimental results of a study of tritium generation and release from the lithium capillary porous system (CPS) under neutron irradiation. The experiments were performed at the IVG1.M research reactor (Kurchatov, Kazakhstan) at a power level of 6MW and temperatures from 325 to 840°С. The studies were performed using two modes: with constant pumping out of the experimental ampoule device (differential mode) and with tritium accumulation in ampoule device (AD) with absence of the pumping out (integrated mode). The diffusion coefficients of free tritium atoms in liquid lithium under neutron irradiation were obtained. The Arrhenius dependence of free tritium atoms diffusion coefficient in lithium was derived as D=0.17·exp(−145/RT).

Effects of different operating conditions on the VOCs removal performance of UV irradiation reactors

Effects of different operating conditions on the VOCs removal performance of UV irradiation reactors

On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

Equilibrium and thermodynamic parameters for heterogeneous esterification of butyric acid with methanol under microwave irradiation

Synthesis of methyl butyrate was investigated in a microwave irradiated batch reactor in presence of acid ion-exchange resin catalyst,amberlyst-15. Methyl ester was heterogeneously produced by the reaction between butyric acid and methanol. Effect of reaction parameters oftemperature (323–343 K), catalyst loading (0–10.5% w/w), alcohol to acid ratio, M (1–5), and amount of molecular sieves added (0–13.5% w/w)on conversion were studied. Equilibrium conversion of 92.6% was achieved in 60 minutes under microwave irradiation. Equilibrium constants atvaried temperatures and dependency of equilibrium constant on temperature were studied. Equilibrium constant and equilibrium conversionshowed increase with the increase in temperature as expected as per le-Chatelier principle. Van’t Hoff plot for esterification of butyric acid waslinear with negative slope indicating that reaction was endothermic. Comparative study showed that microwave irradiated method for methylbutyrate synthesis to be very efficient and fast compared with conventional and ultrasound assisted routes under optimized reaction conditions.

The effect of matrix chemistry on dislocation evolution in an irradiated Zr alloy

Advancements in transmission electron microscopy allow us to draw correlations between evolving matrix chemistry environments and the resulting dislocation structures that form. Such phenomena are essential in predicting the lifetime of neutron reactor components, but are not well understood at the fundamental level. We investigate the effect of nano-scale matrix chemical evolution in Zircaloy-2 on dislocation formation after emulating commercial reactor irradiation conditions on a proton beamline. Similarity in the dislocation type, morphology, density and evolution between the different irradiation types establishes proton irradiation in this regard. For the first time, we observe chemical segregation of Fe, Ni and Cr to a-loop positions in basal traces and the segregation of Sn in alternate rows, anticorrelated to the positions of the light transition elements. The resulting layered structure with a periodicity of ∼50 nm creates an even greater anisotropy than that usually associated with HCP materials. Concurrent analysis of chemical effects and dislocation spatial relationships provides evidence that may explain the delayed onset of c-loop nucleation and accelerated dimensional instability regimes in its dependence on the alignment of a-loops parallel to the trace of the basal plane. This demonstrates the applicability of chemical-structural correlations towards key research questions regarding deformation behaviour.

Investigation of 99Mo potential production via UO2SO4 liquid target irradiation in a 5 MW nuclear research reactor

Abstract The activation method for 99Mo production in comparison to fissionable target irradiation in research reactors is less preferable. Therefore, 99Mo yield using UO2SO4 samples was theoretically investigated. Computational results revealed admirable potential of the liquid samples for 99Mo production. Low-concentrated uranyl sulphate samples could easily be handled by the irradiation box. The sample geometry optimization improves thermal hydraulic conditions and production yield. The optimized geometry including only 0.12 g 235U produced 57Ci99Mo at end-of-irradiation (EOI) with a temperature peak of 72°C during the irradiation.

Evaluation of the Pool Critical Assembly Benchmark with Explicitly Modeled Geometry Using MCNP6

Despite being one of the most widely used benchmarks for qualifying light water reactor (LWR) radiation transport methods and data, no benchmark calculation of the Oak Ridge National Laboratory (ORNL) pool critical assembly (PCA) pressure vessel wall benchmark facility (PVWBF) using MCNP6 with explicitly modeled core geometry exists. As such, this paper provides results for such an analysis. First, a criticality calculation is used to construct the fixed source term. Next, ADVANTG-generated variance reduction parameters are used within the final MCNP6 fixed source calculations. These calculations provide unadjusted dosimetry results using three sets of dosimetry reaction cross sections of varying ages (those packaged with MCNP6, from the IRDF-2002 multigroup library, and from the ACE-formatted IRDFF v1.05 library). These results are then compared to two different sets of measured reaction rates. The comparison agrees in an overall sense within 2% and on a specific reaction and dosimetry location basis within 5%. Except for the neptunium dosimetry, the individual foil raw calculation-to-experiment comparisons usually agree within 10% but are typically greater than unity. Finally, in the course of developing these calculations, geometry that has previously not been completely specified is provided herein for the convenience of future analysts.

A Study of The Assessment for Boron Neutron Capture Therapy Agent

A study of the assessment criteria covers the synthesis and characterization of agent and test their biological effectiveness as boron neutron capture therapy (BNCT) agents in cancer treatment. The cellular uptake of this agent into the glioblastoma cells was assessed by boron analysis (ICP-MS) and by fluorescence imaging (confocal microscopy). The agent enters the glioblastoma cells exhibiting a similar profile, i.e., preferential accumulation in the cytoskeleton and membranes and a low cytotoxic activity (IC<sub>50</sub> values higher than 200 μM). The cytotoxic activity and cellular morphological alterations after neutron irradiation in the Research Reactor (>10<sup>7</sup> neutrons cm<sup>−2</sup> s<sup>−1</sup>) were assessed by the MTT assay and by electron microscopy (TEM). Post neutron irradiation revealed that BNCT has a higher cytotoxic effect on the glioblastoma cells. Results provide a strong rationale for considering one of these compounds as a lead candidate for a new BNCT agent.

Development of a low-level 39Ar calibration standard – Analysis by absolute gas counting measurements augmented with simulation

This paper describes the generation of 39Ar, via reactor irradiation of potassium carbonate, followed by quantitative analysis (length-compensated proportional counting) to yield two calibration standards that are respectively 50 and 3 times atmospheric background levels. Measurements were performed in Pacific Northwest National Laboratory's shallow underground counting laboratory studying the effect of gas density on beta-transport; these results are compared with simulation. The total expanded uncertainty of the specific activity for the ~50× 39Ar in P10 standard is 3.6% (k=2).

Strong trapping and slow diffusion of helium in a tungsten grain boundary

We have investigated the segregation, trapping and diffusion of He in a ∑3<110>{111} W grain boundary (GB) using combined techniques of ab initio and classical atomistic simulations. We show that, with an average segregation energy of −3.20 eV, the strong He trapping can be attributed to a GB interstitial trapping or a vacancy trapping mechanism, while an average energy barrier of 1.97 eV leads to a slow diffusion of He in the GB plane. We further reveal by molecular dynamics simulations that the He diffusion will be dictated by GB migration through the motion of GB disconnections. Interestingly, we also observe a He-induced GB structural transition in classical simulations. The present work suggests that the GB does not provide fast transport channel for He, providing useful reference for the possible application of polycrystalline W under He irradiation in advanced nuclear fusion reactors.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

Mass-Spectrometric Studies of Irradiated Uranium-Erbium Fuel

The aim of this work was to investigate by means of secondary-ion mass spectrometry the isotopic composition of the matrix and the erbium consumable absorber in RBMK fuel samples in the initial state (enrichment from 2.4 to 2.8% in terms of 235U) and the state after irradiation in the IVV-2M reactor (to maximum burnup 34.4 MW·days/kg). It is found that burnup growth is accompanied by reduction of the 167Er concentration and growth of the 168Er content. The residual concentration of 167Er in a mixture of erbium isotopes is 11-fold lower than the initial value. An increase of burnup from 0.42 to 34.4 MW·days/kg is accompanied by 236U and 239Pu accumulation. The 235U concentration in fuel with maximum burnup decreases approximately six-fold relative to the initial value. The usual fission products of irradiated fuel – Cs, Ba, Pr, Nd, Sm, and Ce – are found in the range 132–162 amu.

Equilibrium and thermodynamic parameters for heterogeneous esterification of butyric acid with methanol under microwave irradiation

Abstract Synthesis of methyl butyrate was investigated in a microwave irradiated batch reactor in presence of acid ion-exchange resin catalyst, amberlyst-15. Methyl ester was heterogeneously produced by the reaction between butyric acid and methanol. Effect of reaction parameters of temperature (323–343 K), catalyst loading (0–10.5% w/w), alcohol to acid ratio, M (1–5), and amount of molecular sieves added (0–13.5% w/w) on conversion were studied. Equilibrium conversion of 92.6% was achieved in 60 minutes under microwave irradiation. Equilibrium constants at varied temperatures and dependency of equilibrium constant on temperature were studied. Equilibrium constant and equilibrium conversion showed increase with the increase in temperature as expected as per le-Chatelier principle. Van't Hoff plot for esterification of butyric acid was linear with negative slope indicating that reaction was endothermic. Comparative study showed that microwave irradiated method for methyl butyrate synthesis to be very efficient and fast compared with conventional and ultrasound assisted routes under optimized reaction conditions.

Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors

ABSTRACTMeasured isotopic compositions of UO2 and MOX fuel samples taken from irradiated light water reactor fuel assemblies were analyzed by CASMO5 coupled with a JENDL-4.0 base library to assess the uncertainties in the calculated isotopic compositions on heavy and fission product nuclides. The burnup calculations for the analysis were performed based on a single-assembly model taking into account the detail fuel assembly specifications and irradiation histories. For the MOX fuel samples, a multiple-assembly model was also adopted taking into account the effect of the surrounding UO2 fuel assemblies. The average and standard deviation of the biases (C/E − 1's (here C and E are calculated and measured results, respectively)) were calculated for each nuclide separately on the PWR and BWR UO2 fuel samples. The averaged biases for 235U, 236U, 239Pu, 240Pu, 241Pu and 242Pu were 2.7%, −0.9%, 0.3%, 0.7%, −2.4% and −1.7% for PWR UO2 samples, and 6.7%, −1.5%, 2.5%, −0.6%, 0.4% and −0.1% for BWR UO2 samples, respectively. The biases with the single-assembly model on the MOX fuel samples showed large positive values of 239Pu, and application of the multiple-assembly model reduced the biases as reported in our previous studies.