Effects of Neutron Irradiation on the Structural Performance of the Concrete Biological Shield in Light Water Reactors
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Le Pape, Yann↗
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Results for “irradiation effect”
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Preliminary Investigation of the Effects of Neutron Irradiation on the Weibull Modulus of Graphite
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Campbell, Anne↗
Effect of neutron beamline irradiation on hydrogen equilibrium vapor pressure over hydrided Zircaloy-4
Here, the effect of neutron irradiation on the hydrogen equilibrium vapor pressure over hydrided samples of nickel-plated Zircaloy-4 was evaluated at the Washington State University Research Reactor’s epithermal beamline facility. To enable this study, a measurement capability was designed and assembled to obtain low pressure (<0.1 Pa) measurements over the temperature range of interest (i.e., ~300° - 400°C). Measurement results were in good agreement with tends extrapolated from previous data obtained at higher temperatures as well as recent thermodynamic data. No irradiation enhancement of vapor pressure was observed under the incident flux of epithermal neutrons over this temperature range.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Irradiation-induced shift in the thermodynamic stability of phases and the self-healing effect in transformative high entropy alloys
Here, this work investigated Fe 40 Mn 20 Cr 15 Co 20 Si 5 high entropy alloy (CS-HEA), which exhibits transformation induced plasticity (TRIP) from γ-fcc → ε-hcp, as a probable candidate for nuclear applications. CS-HEA is an extensively-explored, low stacking fault energy alloy with superior strength, ductility, fatigue resistance, and corrosion resistance. This study delved into the effect of irradiation on the shift in thermodynamic stability of the phases and thus radiation tolerance. The evolution of phases, lattice parameters, and transformation volume, V γ→ε , were evaluated from X-ray diffraction experiments along with the mechanical response from nanoindentation. The alloy exhibited a recently-proposed novel self-healing mechanism possible due to the TRIP effect to minimize irradiation damage by restraining the γ-fcc → ε-hcp transformation via thermal aid; this self-healing mechanism was confirmed by transmission electron microscopy. The results were corroborated by a negative change in V γ→ε and a low |V γ→ε |, which is an important criterion for recovery of parent γ-fcc phase. Thus, this alloy was deemed a good radiation-tolerant candidate for nuclear application.
36 MATERIALS SCIENCE↗
4D evolution of Cr 23 C 6 precipitates in neutron-irradiated and annealed HT-UPS steel observed via synchrotron micro-computed tomography
High-temperature-ultrafine precipitate strengthened (HT-UPS) steel is a potential structural material for advanced nuclear reactors; however, its irradiation response is not well understood. This research provides insight into irradiation-induced effects, such as precipitate evolution mechanisms and four-dimensional morphological evolution, in HT-UPS steel using synchrotron micro-computed tomography. Identical specimens were characterized pre-irradiation and post-irradiation following neutron exposure up to 0.3 displacements per atom at 600 °C. Irradiation effects were also differentiated from the annealing response of precipitates. Following neutron irradiation, the average Cr 23 C 6 precipitate size reduced, affected by the synergy of nucleation and growth, ballistic dissolution, and inverse coarsening, which was observed at fluences an order of magnitude lower than previously observed. Annealing at 600 °C for 32 h increased the average Cr 23 C 6 precipitate size and decreased the phase fraction, attributed to precipitate coarsening. The precipitate morphology evolution and resultant mechanisms can be utilized to parameterize and validate microstructural models simulating radiation damage or annealing.
36 MATERIALS SCIENCE↗
Effect of initial γ -irradiation on infrared laser ablation of poly(vinyl alcohol) studied by infrared spectroscopy
We report the effect of initial γ-irradiation of poly(vinyl alcohol) (PVA) on the kinetics of polymer ablation under CO 2 laser irradiation was studied using infra-red spectroscopy. The rate of laser ablation rapidly decreases with increasing time of laser ablation up to 10 s. Initial γ-irradiation of a dose up to 100 kGy reduces the rate of the laser ablation, just like the effect of laser exposure. The number of OH and CH groups decrease in PVA both with laser irradiation and with γ-irradiation. The number of OH groups decreases more under laser irradiation. Unsaturated bonds are formed in γ–irradiated PVA on the surface of the crater formed by laser ablation of the polymer that has undergone radiolysis. Increasing the dose of initial γ-irradiation from 100 to 2300 kGy leads to an increase in the intensity of the IR bands of the unsaturated bonds. The number of IR bands in the spectra of the laser ablation crater of PVA does not change with the dose of initial γ-irradiation. A dose of radiolysis of 3500 kGy significantly changes the IR spectra of both the ablation crater and the coating formed from the ablation products. Many bands characteristic for PVA are no longer present and the intensities of bands corresponding to unsaturated bonds are increased. A computational model for loss of water and ketone formation was developed for PVA and PVA including a diol using G3(MP2). The highest activation energy barrier step in the decomposition reaction was formation of the enol species and water was shown to catalyze all decomposition steps.
36 MATERIALS SCIENCE↗
IASCC of 304 SS in BWR environments: Effects of post-irradiation annealing and surface condition
To investigate the impact of low-temperature post-irradiation annealing (PIA) on the stress corrosion cracking of neutron-irradiated 304 stainless steel, constant extension rate tests were conducted in simulated boiling water reactor normal water chemistry (BWR-NWC) and hydrogenated water chemistry (HWC) environments. Ten tensile samples, comprising five as-irradiated and five PIA specimens, were prepared by electropolishing the gauge section of electric discharge machined (EDM) samples. Here, the annealing treatment reduced the yield strength from approximately 550 MPa to around 425 MPa and significantly restored the ductility and the strain hardening capability of the alloy. Consequently, the susceptibility of this material to irradiation-assisted stress corrosion cracking (IASCC) was effectively mitigated, which is more prominent in HWC, as evident from fractography, which indicated a decreased propensity for intergranular (IG) fracture. Furthermore, it was observed that the polished surface facilitated crack initiation more readily than the EDM surface, suggesting that the EDM process suppressed crack initiation to some extent.
Crack initiation↗
In-Situ Ion Irradiation to Add Irradiation Assisted Grain Growth to the Marmot Tool
In the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the MARMOT mesoscale fuel performance tool is used to inform the development of mechanistic materials models for the BISON fuel performance tool. The grain size of the fuel has a large impact on its performance, directly impacting heat conduction, fission gas release, creep, and fracture. Thus, atomistic and mesoscale MARMOT simulations have been used to investigate grain boundary migration and grain growth in UO 2 . However, the current grain growth models in MARMOT does not consider the effect of irradiation on causing grain growth. This research project experimentally studied and implemented irradiation effects on grain growth of UO 2 into MARMOT.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The effect of gamma ray irradiation on few layered MoSe2: A material for nuclear and space applications
In recent years, emerging two-dimensional (2D) materials, such as molybdenum diselenide (MoSe2), have been at the center of attention for many researchers. This is due to their unique and fascinating physicochemical properties that make them attractive in space and defense applications that include shielding harsh irradiation environments. In this study, we examined the effects of gamma (γ) rays at various doses on the structural, chemical, and optical properties of MoSe2 layers. After the samples were exposed to intense gamma radiation (from a 60Co source) with various exposure times to vary the total accumulated dosage (up to 100 kGy), Raman and photoluminescence spectroscopies were used to study and probe radiation-induced changes to the samples. When compared to pristine materials, very few changes in optical properties were typically observed, indicating good robustness with little sensitivity, even at relatively high doses of gamma radiation. The imaging using scanning electron microscopy revealed a number of nano-hillocks that were connected to substrate alterations. X-ray photoelectron spectroscopies revealed that Mo’s binding energies remained the same, but Se’s binding energies blueshifted. We associated this shift with the decrease in Se vacancies that occurred after irradiation as a result of Mo atoms creating adatoms next to Se atoms. When compared to pristine materials, very few changes in optical, chemical, and structural properties were typically observed. These findings highlight the inherent resilience of MoSe2 in hostile radioactive conditions, which spurs additional research into their optical, electrical, and structural characteristics as well as exploration for potential space, energy, and defense applications.
Materials Science↗
Investigation of Nano-Scale Segregation in Nanostructured Ferritic Alloy 14YWT after Heavy Ion Irradiation
Oxide-dispersion-strengthened (ODS) steels, which contain nano-scale Y-Ti-O particles, are being considered for high-temperature radiation environments of nuclear reactors. It is important to accurately characterize the structure of grain boundaries and understand the behavior of segregation at grain boundaries in ODS steels during irradiation. The effect of heavy ion irradiation at 700 °C on Nanostructured Ferritic Alloy 14YWT was investigated using Atom Probe Tomography. Enrichment of Cr occurs at the grain boundaries as well as at nano oxide particle surfaces in the unirradiated sample. The enrichment of Ti and Y at a grain boundary corresponds with Y-Ti-O nano oxide particles with larger size compared to those in the grain, and the Cr enrichment is particularly accentuated at these larger nano oxide particles. The segregation of W occurs at the grain boundaries that are without nano oxide particles. O is segregated at grain boundaries without oxide particles after irradiation. The segregation behavior of Cr, W, Ti, and Y at the grain boundary in the irradiated samples is similar to that in the unirradiated sample. The nano oxide particles embedded in the grain boundary are a primary reason for the increase in Cr segregation at the grain boundary.
36 MATERIALS SCIENCE↗
Effects of γ-Irradiation, Cation Size, and Salt Concentration on the Aggregation of Boehmite Nanoplatelets
The effects of radiation on the aggregation of nanoplatelets of aluminum oxyhydroxide (boehmite) in slurries of M 1+ cation nitrates has been observed using tumbler small- and ultra-small-angle neutron scattering. Nitrate solutions of H, Li, Na, K, and Rb at concentrations of 10 -5 , 10 -3 , 10 -1 , 2, and 4 molal were compared to the results for pure H 2 O. Primary aggregates consisting of small stacks of boehmite platelets that form immediately after the boehmite is placed in water are smaller for irradiated boehmite than the pristine material. Here, the primary aggregates for irradiated boehmite also appear to have a rougher surface than those observed for unirradiated materials. Both results are possibly due to breaking of the surface OH species and a neutralization of the surface. At low salt concentrations, the pH of the irradiated boehmite slurries is lower than that of the corresponding pristine material indicating that irradiated boehmite adsorbs fewer protons or more hydroxide onto its surface than the pristine boehmite. Increasing the salt concentration decreases the pH to near neutral at 7 due to screening of the electric double layer at the solid/liquid interface. Irradiation does not seem to affect the size of secondary aggregates formed from the aggregation of primary aggregates or of the overall size of the primary aggregates.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Design Report on the MiniFuel Instrumented Test Apparatus for Understanding Radiation Effects
Most nuclear fuels irradiations at Oak Ridge National Laboratory (ORNL) over the past decade have been conducted using MiniFuel—a static capsule design employing subscale fuel specimens to collect separate-effects irradiated fuel performance data. Irradiation conditions for MiniFuel experiments are predicted pre-test using reactor physics, and thermal models are verified post-irradiation via SiC dilatometry and various spectrometry methods. Relevant fuel performance parameters are also observed post-irradiation in a hot cell, thereby providing a single data point for each parameter representing the cumulative effects of the irradiation conditions. Substantially more data can be harvested from a single test and within a shorter duration by instrumenting irradiation vehicles and measuring desired quantities in situ. This report presents the design and analysis of the MiniFuel INstrumented Irradiation Test Apparatus for Understanding Radiation Effects (MINITAURE)—an instrumented test rig based on the separate-effects MiniFuel concept that aims to capture fission gas release (FGR) and thermal conductivity degradation of fuel specimens during irradiation in the High Flux Isotope Reactor (HFIR). MINITAURE will be integrated with the Materials Irradiation Facility (MIF) located in the HFIR building outside the reactor containment. The MIF will act as the instrumentation and control center for the experiment, enabling real-time feedback from in situ sensors and control of irradiation temperatures via a gas delivery system. Two unique capsule designs were developed to capture each phenomenon: the thermal conductivity capsule, which uses a thermopile method to estimate fuel specimen thermal conductivity, and the fission gas release capsule, which will have continuous flowing gas communication to high-purity germanium detectors that are housed in the MIF for monitoring FGR. This report details the reactor physics and heat transfer modeling activities that were used to inform the experiment design and predict capsule performance. It also describes out-of-pile activities conducted to stand up this new capability and verify the measurement techniques. Modeling efforts to date have demonstrated the feasibility of the in situ measurement techniques and supported the development of the MINITAURE assembly configuration. Out-of-pile testing of the thermal conductivity measurement shows promise in capturing relative changes in thermal conductivity. However, significant errors exist in the measured absolute value, posing a need for further refinement.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Photoluminescence and Raman spectroscopy of wide bandgap semiconductors damaged by deep-UV laser irradiation
The effects of a pulsed, focused, deep-UV (4.66 eV) laser on wide and ultra-wide bandgap semiconductors were investigated with photoluminescence (PL) and Raman spectroscopy. Three semiconductor single crystals were studied: silicon carbide (6H-SiC), gallium nitride (GaN), and gallium oxide (β-Ga 2 O 3 ). Atomic emission lines from neutral Ga or Si were observed during the laser-damage process. For all three semiconductors, PL mapping (3.49 eV laser excitation) of the damaged material revealed visible emission bands in the 2.6–2.8 eV range, attributed to point defects. Raman spectra (2.33 eV excitation) showed a reduction in the Raman peak intensities in the damaged region, along with weak PL bands around 1.9–2.1 eV.
36 MATERIALS SCIENCE↗
Raman spectroscopy of neutron irradiated silicon carbide
The effects of neutron irradiation on microstructural evolution and the resultant changes in physical and mechanical properties are of critical importance for the development of silicon carbide (SiC) materials for nuclear applications. This study neutron-irradiated βSiC under a wide range of conditions at temperatures between 235 and 750°C and neutron doses of 0.01–11.8 displacements per atom, and then evaluated the effects on the SiC structure using Raman spectroscopy. The SiC optical phonon lines were shifted to lower wavenumbers by irradiation. Correlations were found among the wavenumber of the longitudinal optical phonon line, irradiation-induced swelling, and irradiation temperature. The peak shift also correlated indirectly with decreasing thermal conductivity of irradiated SiC. The irradiation-induced peak shift is explained by combinations of lattice strain, reduction of the elastic modulus, and other factors including decreasing coherent domain size. These findings bridge irradiation-induced microstructural changes and property changes and illustrate how Raman spectroscopy is a useful tool for nondestructively assessing irradiated SiC materials for nuclear applications.
Koyanagi, T.↗
Effects of Radiation-Induced Defects on Corrosion
The next generation of nuclear reactors will expose materials to conditions that, in some cases, are even more extreme than those in current fission reactors, inevitably leading to new materials science challenges. Radiation-induced damage and corrosion are two key phenomena that must be understood both independently and synergistically, but their interactions are often convoluted. In the light water reactor community, a tremendous amount of work has been done to illuminate irradiation-corrosion effects, and similar efforts are under way for heavy liquid metal and molten salt environments. While certain effects, such as radiolysis and irradiation-assisted stress corrosion cracking, are reasonably well established, the basic science of how irradiation-induced defects in the base material and the corrosion layer influence the corrosion process still presents many unanswered questions. In this review, we summarize the work that has been done to understand these coupled extremes, highlight the complex nature of this problem, and identify key knowledge gaps.
36 MATERIALS SCIENCE↗
Structural and thermodynamic evolution of an amorphous SiOC ceramic after swift heavy ion irradiation
Radiation induced structural and energetic changes have been widely studied in crystalline materials, but studied to a much lesser extent in amorphous solids. Using a combination of neutron pair distribution function (PDF) analysis and high temperature oxide melt solution calorimetry, we obtained critical insights into the structural and energetic evolution in a swift heavy ion irradiated amorphous SiOC polymer-derived ceramic. The radiation modified SiOC structure becomes energetically less stable by 24.0±1.6 kJ/mol compared to the unirradiated structure. This destabilization is related to decreased connectivity of the Si-O-C network and destruction of free carbon. In comparison with unirradiated SiOC, the irradiated structure is more likely to become phase separated during subsequent thermal annealing. Here, our study has important implications for evaluating amorphous SiOC ceramic as a possible radiation resistant structure for nuclear applications.
36 MATERIALS SCIENCE↗
UNCERTAINTIES FOR IN-REACTOR TESTING OF TRITIUM LOADED GETTERS WITH TEMPERATURE CONTROL
The fourth experiment in TPBAR Materials Irradiation Separate-Effects Test (TMIST) series (TMIST-4) is planned to experimentally determine the effect of neutron irradiation on the equilibrium vapor pressure of tritium over a tritium loaded Nickel-Plated Zirconium (NPZ) getter in a simulated TPBAR service environment. This pressure will be used to estimate the tritium permeation rate. In the thermal design in the TMIST-4, two evaluations are required. 1) The cladding temperature is calculated to estimate the feasibility of detecting potential irradiation enhancement of the partial pressure for an NPZ getter. This work is required to obtain a precise cladding temperature for measurement of a low tritium partial pressure. This work has been completed and indicated that the required/target cladding temperatures based on the TMIST-4 test plan can be satisfied under the test condition range by modifying the capsule component shapes and inserting a thermally conductive metal sleeve. 2) The evaluation of uncertainties associated with the test design and operation is performed. This includes fabrication tolerances, and power and nuclear operating conditions. These factors will impact the component temperatures inside capsules and resulting the tritium partial pressure. The tritium permeation rate is a function of temperature and tritium partial pressure. The prediction of tritium permeation rates affects the feasibility of obtaining measurement resolutions required to reach reasonable conclusion for potential irradiation enhanced performance. This work will investigate the uncertainties.
TTP, TPBAR, TMIST-4, TPBAR Materials Irradiation S↗
Effects of helium on irradiation response of reduced-activation ferritic-martensitic steels: Using nickel isotopes to simulate fusion neutron response
Understanding the effects of helium on microstructures and mechanical properties of reduced-activation ferritic-martensitic steels is important to use of these steels in fusion reactor structures. The 9Cr-2WVTa steels were doped with 58Ni and 60Ni isotopes at 2 weight percent to control the rate of transmutation helium generation. The samples were irradiated in the High Flux Isotope Reactor to ~24 displacements per atom at nominal temperatures of 300, 400, and 500°C, producing 228 and 7 atomic parts-per-million helium in the 58 Ni- and 60 Ni-doped samples, respectively. Transmission electron microscopy revealed a variety of precipitates and the radiation-induced dislocation loops and cavities (voids or helium bubbles). Tensile tests of the irradiated samples at the irradiation temperatures showed radiation-induced hardening at 300°C and radiation-induced softening at 400°C. Analysis indicates that the hardening primarily originated from the loops and cavities. The 58 Ni-doped samples had greater strengthening contributions from loops and cavities, leading to higher hardening with lower ductility than the 60 Ni-doped samples. The greater helium production of 58 Ni did not show pronounced reductions in ductility of the samples.
36 MATERIALS SCIENCE↗