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At least 487 records · Page 27

Comparison of Thermal Diffusivity Degradation of Irradiated Zirlo and SiC-SiC Composite

Zirlo (Zircaloy cladding material from Westinghouse) has been widely used as the fuel cladding material in Light Water Reactors because of its high melting point, low susceptibility to corrosion, good thermal and mechanical shock resistance, and low degradation of mechanical properties under irradiation. But as Zirlo reacts with water at high temperature, which can trigger serious accidents, efforts have been made to explore better candidates to replace Zirlo in the water-cooled reactors. SiC-SiC composite is considered as one of the most viable options due to its comparable mechanical performances and corrosion resistance. However, one of the most important properties, thermal diffusivities of both Zirlo and SiC-SiC composite, and their degradation after irradiation, are not well known yet. Ion-irradiated samples are used to study the thermal property degradation caused by irradiation instead of the neutron-irradiated ones due to the significantly lower radioactivity. The microstructural damage is estimated from the software simulation SRIM/TRIM based on the energy and fluence of the ion, and the target material, and represented by DPA (displacement by atom). Different from neutron-irradiated samples, the damaged layer of ion-irradiated samples is only a few to a few tens of micrometers thick, and the damage is depth dependent. It is impossible to directly measure the bulk property change, and a microscale measurement technique is necessary. Laser-based photothermal reflectance technique is one of the ideal methods to measure local thermal conductivity and diffusivity with high spatial resolution. The main principle of this technique is to heat the solid sample using an intensity-modulated laser, and scan the sample with a constant intensity continuous-wave (CW) laser to detect the thermal wave propagation. On bulk, isotropic material, if the heating laser intensity is modulated as a sine wave, thermal diffusivity of the sample can be derived from the phase lag of the thermal wave, scan distance and modulation frequency in a straight forward manner. By using more specific analytical models, thermal conductivity and diffusivity in complicated cases, such as layered structure materials, composite materials, and/or anisotropic materials, are measurable as well. In this work, thermal diffusivity of the light damage layer of the post-irradiation Zirlo and SiC-SiC composite are measured and compared. The Zirlo sample is irradiated by Ni{sup 9+} ion, and SiC-SiC composite material is irradiated by Si{sup 7+} ion. The peak DPA of both sample is similar as ∼10. For ion-irradiated materials, DPA is known to increase with depth before it reaches the peak, and then sharply drop to zero. In term of light damage layer in this manuscript, we mean by the layer with a DPA smaller than 5% maximum value. In this layer DPA increases slowly and smoothly with depth, and the structure and thus thermal properties are more spatially uniform. For the SiC-SiC composite, the measurements are separately conducted on fibers along radial or axial direction, and on the matrix. Our measurement results show that with similar microstructural damage, thermal diffusivity of matrix SiC reduces more significantly than that of Zirlo. For fiber SiC that has a much lower thermal diffusivity than the matrix SiC before irradiation, the post-irradiation effect on thermal diffusivity is not obvious.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Fabrication of nuclear reactor components using additive manufacturing (AM) methods is now a practical option since the AM technologies have advanced to allow for building of complex parts with high quality materials. To assess the mechanical performance of printed components in reactor-relevant conditions and to build a property database for the AM 316L stainless steel (SS), mechanical testing and characterization were performed before and after neutron irradiation. In this work, miniature tensile specimens were irradiated at the High Flux Isotope Reactor (HFIR) to 0.2 and 2 displacements per atom (dpa) at 300 and 600°C. The AM 316L SS was tested in the as-built, stress-relieved, and solution-annealed conditions, and the wrought (WT) 316L SS in solution-annealed condition as a reference alloy. The baseline test result showed that the AM 316L SS, regardless of the post-build heat treatment, had higher strength than the WT 316L SS, but similar ductility. Post-irradiation tensile testing was conducted at RT, 300°C, and 500°C for selected irradiation conditions. Neutron irradiation induced significant changes in the mechanical behavior of the AM stainless steels, including both hardening and softening. Although the as-built 316L steel after 300°C irradiation showed necking just after yielding, the overall property changes of the as-printed alloy became less significant after 600°C irradiation. Irradiation-induced ductilization was also observed after the higher temperature irradiation. In general, the strength change was smaller in the relatively stronger as-built and stress-relieved AM SSs than in the solution-annealed AM and WT SSs. These relatively lower strength 316L SSs overall retained higher ductility in the irradiation conditions tested, but the stronger 316L SSs demonstrated a similar level of ductility after the higher temperature (600°C) irradiation. It is a positive assessment for the AM 316L materials that no embrittlement was observed within the test and irradiation conditions of the experiment.

36 MATERIALS SCIENCE↗

Post-irradiation examination of AGR-3/4 TRISO fuel compacts using three-dimensional X-ray computed tomography

The AGR-3/4 irradiation tests combined the third and fourth planned irradiation experiments in the US Department of Energy’s Advanced Gas Reactor (AGR) testing campaign of tri-structural isotropic (TRISO) fuel compacts. In this article, we present post-irradiation examination (PIE) using X-ray computed tomography (XCT) of two unirradiated and two irradiated compacts from the AGR-3/4 irradiation tests. The irradiated compacts studied (compact 7–1 and compact 12–4) represent the upper and lower limit of burnup within the AGR-3/4 irradiation experiment. This article presents a detailed quantitative analysis on the post-irradiation structure of TRISO fuel compacts. Various quantitative parameters including shape, size, and packing of kernels, and their spatial distribution, were utilized to gain insights into the structural changes caused by irradiation. The equivalent diameter and sphericity were found to increase and decrease, respectively, in irradiated compact 7–1 due to its higher burnup. Nearest neighbor distance between fuel kernels decreased after irradiation, suggesting irradiation-induced shrinkage of graphitic matrix. Furthermore, each compact in AGR-3/4 irradiation tests contained 20 designed-to-fail (DTF) fuel particles that were meant to act as a source of fission product release to the experiment test train. Furthermore, in the present work, all DTF fuel particles in the four compacts studied were identified, and it was found that they exhibited larger kernel swelling in compact 12–4 and smaller kernel swelling in compact 7–1, compared to the driver particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Long-term total solar irradiance variability during sunspot cycle 22

Total solar irradiance measurements from the 1984-1993 Earth Radiation Budget Satellite (ERBS) active cavity radiometer and 1978-1993 Nimbus 7 transfer cavity radiometer spacecraft experiments are analyzed to detect the presence of 11-, 22-, and 80-year irradiance variability components. The analyses confirmed the existence of a significant 11-year irradiance variability component, associated with solar magnetic activity and the sunspot cycle. The analyses also suggest the presence of a 22- or 80-year variability component. The earlier Nimbus 7 and Solar Maximum Mission (SMM) spacecraft irradiance measurements decreased approximately 1.2 and 1.3 W/sq m, respectively, between 1980 and 1986. The Nimbus 7 values increased 1.2 W/sq m between 1986 and 1989. The ERBS irradiance measurements increased 1.3 W/sq m during 1986-1989, and then decreased 0.4 W/sq m (at an annual rate of 0.14 W/sq. m/yr) during 1990-1993. Considering the correlations between ERBS, Nimbus 7, and SMM irradiance trends and solar magnetic activity, the total solar irradiance should decrease to minimum levels by 1997 as solar activity decreases to minimum levels, and then increase to maximum levels by the year 2000 as solar activity rises. The ERBS measurements yielded 165.4 +/- 0.7 W/sq m as the mean irradiance value with measurement accuracies and precisions of 0.2% and 0.02%, respectively. The ERBS mean irradiance value is within 0.2% of the 1367.4, 1365.9, and 1366.9 W/sq m mean values for the SMM, Upper Atmosphere Research Satellite (UARS), and Space Shuttle Atmospheric Laboratory for Applications and Science (ATLAS 1) Solar Constant (SOLCON) active cavity radiometer spacecraft experiments, respectively. The Nimbus 7 measurements yielded 1372.1 W/sq m as the mean value with a measurement accuracy of 0.5%. Empirical irradiance model fits, based upon 10.7 -cm solar radio flux (F10) and photometric sunspot index (PSI), were used to assess the quality of the ERBS, Numbus 7, SMM, and the UARS irradiance data sets and to identify irradiance variability trends which may be caused by drifts or shifts in the spacecraft sensor responses. Comparisons among the fits and measured irradiances indicate that the Nimbus 7 radiometer response shifted by a total of 0.8 W/sq m between September 1989 and April 1990 and that the ERBS and UARS radiometers each drifted approximately 0.5 W/sq m during the first 5 months in orbit.

Lee, Robert B., III↗

The effect of in vivo and in vitro irradiation (25 Gy) on the subsequent in vitro growth of satellite cells

The effect of in vivo and in vitro irradiation on subsequent satellite cell growth, in vitro, was investigated to ascertain the ability of a 25 Gy dose to inhibit satellite cell proliferation. Satellite cells were isolated from the left (irradiated) and right (non-irradiated) Pectoralis thoracicus of two-week-old tom turkeys 16 h (n=3) and seven weeks (n=2) after the left Pectoralis thoracicus had been irradiated (25 Gy). Satellite cells isolated from the irradiated and non-irradiated muscles exhibited similar (P>0.10) in vitro proliferation indicating that a population of satellite cells survived an in vivo dose of 25 Gy. In additional experiments, satellite cell cultures derived from tom turkey Pectoralis thoracicus were irradiated (25 Gy) in vitro. The number of satellite cells did not (P>0.05) increase in irradiated cultures for 134 h following irradiation, while satellite cells in non-irradiated cultures proliferated (P<0.05) over this time. At later time periods, satellite cell number increased (P<0.05) in irradiated cultures indicating that a population of satellite cells survived irradiation. The results of these in vitro experiments suggest that a 25 Gy dose of irradiation does not abolish satellite cell divisions in the turkey Pectoralis thoracicus.

NASA Discipline Musculoskeletal↗

Microstructural evolution and transmutation in tungsten under ion and neutron irradiation

This study aims to compare the effects of neutron and self-ion irradiation on the mechanical properties and microstructural evolution in W. Neutron irradiation at the HFR reactor to 1.67 dpa at 800 °C resulted in the formation of large Re and Os rich clusters and voids. The post-irradiation composition was measured using APT and verfified against FISPACT modelling. The measured Re and Os concentration was used to create alloys with equivalent concentrations of Re and Os. These alloys were exposed to self-ion irradiation to a peak dose of 1.7 dpa at 800 °C. APT showed that self-ion irradiation leads to the formation of small Os clusters, wheras under neutron irradiation large Re/Os clusters form. Voids are formed by both ion and neutron irradiation, but the voids formed by neutron irradiation are larger. By comparing the behaviour of W-1.4Re and W-1.4Re-0.1Os, suppression of Re cluster formation was observed. Irradiation hardening was measured using nanoindentation and was found to be 2.7 GPa, after neutron irradiation and 1.6 GPa and 0.6 GPa for the self-ion irradiated W-1.4Re and W-1.4Re-0.1Os. The higher hardening is attributed to the barrier strength of large voids and Re/Os clusters that are observed after neutron irradiation.

36 MATERIALS SCIENCE↗

AGR-1 Irradiation Test Final As-Run Report, Rev. 3

This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 10 25 n/m 2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10 -7 with only one capsule significantly exceeding this value. A maximum R/B of around 2 x 10 -7 was reached at the end of the irradiation in Capsule 5. Several shakedown issues were encountered and resolved during the first three cycles. These include the repair of minor gas line leaks; repair of faulty gas line valves; the need to position moisture monitors in regions of low radiation fields for proper functioning; the enforcement of proper on-line data storage and backup, the need to monitor thermocouple performance, correcting for detector spectral gain shift, and a change in the mass flow rate range of the neon flow controllers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Investigation of the mechanical and microstructural evolution of a Cu based bulk metallic glass during ion irradiation

Ion irradiation and annealing experiments were performed on Cu 60 Zr 20 Hf 10 Ti 10 bulk metallic glass (BMG) specimens to investigate their irradiation- and temperature-induced microstructural and mechanical property evolution. For the ion irradiations, samples were exposed to 9 MeV Ni 3+ ions to a midrange (~1.2 μm depth) dose of 10 displacements per atom (dpa) at temperatures ranging from room temperature to 360 °C (the corresponding peak dose at ~2.8 μm depth was ~25 dpa). Bulk X-ray diffraction (XRD) and transmission electron microscopy (TEM) characterization revealed that the alloy did not crystallize during irradiation up to 290 °C but did partially crystallize at 360 °C. XRD analysis revealed that the crystallization which occurred in the sample irradiated at 360 °C was caused by thermal effects instead of irradiation displacement damage. Subsequent Rietveld refinement analysis of the XRD measurements revealed the presence of two distinct crystal phases, namely a CuTiZr hexagonal structure belonging to the P63/mmc space group and a CuTi tetragonal structure belonging to the P4/mmm space group. Nanoindentation experiments revealed that no pronounced hardness changes occurred in the specimens irradiated at room temperature and 290°C, although significant hardening was observed in the sample irradiated at 360 °C. The significant increase in the hardness at 360°C was ascribed to thermally induced partial crystallization of the alloy instead of the ion irradiation. In general, the results of the nanoindentation experiments and XRD characterization suggest that although the Cu BMG exhibits good stability during irradiation at temperatures up to 290 °C it is not suitable for irradiation environments where the temperature is 360 °C for extended periods of time. The Lam and Chong extrapolation method, which has been used to study the indentation size effect (ISE) in amorphous alloys, was employed to quantify how irradiation and temperature affect this type of behavior in the BMG. However, the poor linear fitting of the indentation hardness data by this model indicate that a new ISE model is likely needed to quantify indentation hardening in BMGs.

36 MATERIALS SCIENCE↗

Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux

The purpose of this work is to design an irradiation vehicle for testing silicon carbide (SiC) fiber-reinforced SiC matrix composite cladding materials under conditions representative of a light water reactor in order to validate thermo-mechanical models of stress states in these materials due to irradiation swelling and differential thermal expansion. The design allows for a constant tube outer surface temperature in the range of 300–350 °C under a representative high heat flux (~0.66 MW/m 2 ) during one cycle of irradiation in an un-instrumented “rabbit” capsule in the High Flux Isotope Reactor. An engineered aluminum foil was developed to absorb the expansion of the cladding tubes, due to irradiation swelling, without changing the thermal resistance of the gap between the cladding and irradiation capsule. Finite-element analyses of the capsule were performed, and the models used to calculate thermal contact resistance were validated by out-of-pile testing and post-irradiation examination of the foils and passive SiC thermometry. Six irradiated cladding tubes (both monoliths and composites) were irradiated and subsequently disassembled in a hot cell. The calculated temperatures of passive SiC thermometry inside the capsules showed good agreement with temperatures measured post-irradiation, with two calculated temperatures falling within 10 °C of experimental measurements. Furthermore, the success of this design could lead to new opportunities for irradiation applications with materials that suffer from irradiation swelling, creep, or other dimensional changes that can affect the specimen temperature during irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A novel in-situ, lift-out, three-point bend technique to quantify the mechanical properties of an ex-service neutron irradiated inconel X-750 component

A first of its kind small scale mechanical testing technique involving micro-three-point bending was invented, developed, and implemented on reactor irradiated, active Inconel X-750 components removed from service after approximately 53 and 67 dpa. These tests were performed at ambient room temperature in-situ using a scanning electron microscope in order to obtain live recordings of sample deformation and loading curves. Sample and testing apparatus preparation required novel lift-out and fabrication processes. Materials from two irradiation temperature regimes, low temperature (120-280 °C) and high temperature (300 ± 15 °C) were examined. Manufacturing and finishing (grinding) of this component create differences between its edge and center, so micro-specimens from both areas were extracted in order to study these differences. According to three-point beam bending theory, a 0.2% offset yield stress parameter is introduced and calculated for all specimens. Differences in mechanical properties due to irradiation temperature and dose effects were observed. Material irradiated at the higher temperature exhibited yield strength increases of ~540 MPa after 53 dpa and ~1000 MPa after 67 dpa. There was little difference (≤310 MPa) in yield strength between materials irradiated at the lower temperature at 53 dpa and 67 dpa compared with non-irradiated material. Differences in yield strengths between the edge and center of the component are retained after irradiation. The difference in yield strengths for the edge and center regions was ~740 MPa for non-irradiated material. After irradiation to a dose of 67 dpa these differences were ~570 MPa for the lower irradiation temperature and ~710 MPa for higher irradiation temperature. There were no indications of grain boundary failures via cracking except for material irradiated to 67 dpa at low temperature.

36 MATERIALS SCIENCE↗

Irradiation creep measurement and microstructural analysis of chromium nitride–coated zirconium alloy using pressurized tubes

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. Here, to accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300°C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8 $\times$ 10 21 n/cm 2 fast fluence (E n > 10 MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

36 MATERIALS SCIENCE↗

SU-E-T-266: Development of Evaluation System of Optimal Synchrotron Controlling Parameter for Spot Scanning Proton Therapy with Multiple Gate Irradiations in One Operation Cycle

Purpose: We have developed a gated spot scanning proton beam therapy system with real-time tumor-tracking. This system has the ability of multiple-gated irradiation in a single synchrotron operation cycle controlling the wait-time for consecutive gate signals during a flat-top phase so that the decrease in irradiation efficiency induced by irregular variation of gate signal is reduced. Our previous studies have shown that a 200 ms wait-time is appropriate to increase the average irradiation efficiency, but the optimal wait-time can vary patient by patient and day by day. In this research, we have developed an evaluation system of the optimal wait-time in each irradiation based on the log data of the real-time-image gated proton beam therapy (RGPT) system. Methods: The developed system consists of logger for operation of RGPT system and software for evaluation of optimal wait-time. The logger records timing of gate on/off, timing and the dose of delivered beam spots, beam energy and timing of X-ray irradiation. The evaluation software calculates irradiation time in the case of different wait-time by simulating the multiple-gated irradiation operation using several timing information. Actual data preserved in the log data are used for gate on and off time, spot irradiation time, and time moving to the next spot. Design values are used for the acceleration and deceleration times. We applied this system to a patient treated with the RGPT system. Results: The evaluation system found the optimal wait-time of 390 ms that reduced the irradiation time by about 10 %. The irradiation time with actual wait-time used in treatment was reproduced with accuracy of 0.2 ms. Conclusion: For spot scanning proton therapy system with multiple-gated irradiation in one synchrotron operation cycle, an evaluation system of the optimal wait-time in each irradiation based on log data has been developed. Funding Support: Japan Society for the Promotion of Science (JSPS) through the FIRST Program.

60 APPLIED LIFE SCIENCES↗

Effects of irradiation temperature on the microstructure and deuterium retention in γ-LiAlO 2 pellets

This report presents the experimental findings obtained from November 2023 to September 2024. The study aims to investigate the effects of temperature on the microstructure, deuterium (D) retention, and lithium (Li) loss in γ-LiAlO 2 pellets subjected to sequential He + and D + ion irradiation to a high dose. Sequential ion irradiation has been a key method in our previous studies to simulate the behavior of γ-LiAlO 2 under neutron irradiation. In addition to dose and dose rate, irradiation temperature is a critical factor influencing microstructural and compositional changes. LiAl 5 O 8 precipitates have been observed in γ-LiAlO 2 pellets irradiated with reactor neutrons at ~300°C. These precipitates also form during ion irradiation at an elevated temperature of 500°C, but not at 300°C. In our ion irradiation experiments, the dose rate is typically three orders of magnitude higher than that of neutron irradiation, leading to a more rapid damage production. To better emulate the microstructural features in neutron-irradiated pellets using ion irradiation, a higher irradiation temperature is needed to accelerate the diffusion of point defects and enhance defect recovery rates, thereby compensating for the effects of the higher dose rate. The microstructural changes observed are the result of competing processes occurring during ion irradiation.

36 MATERIALS SCIENCE↗

Nuclear fuel irradiation testbed for nuclear security applications

The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Microstructure, electrical resistivity, and tensile properties of neutron-irradiated Cu–Cr–Nb–Zr

High strength, high conductivity copper alloys that can resist creep at high temperatures are one of the primary candidates for efficient heat exchangers in fusion reactors. Cu–Cr–Nb–Zr (CCNZ) alloys, which were designed to improve the strength and creep life of ITER Cu–Cr–Zr (CCZ) reference alloys, have been found to have comparable electrical conductivity and tensile properties to CCZ alloys. The measured creep rupture times for these improved alloys is about ten times higher than the ITER reference alloys at 90–125 MPa at 500 °C. However, the effects of neutron irradiation on these alloys, and the ensuing material properties, have not been studied; thus, their utility in a fusion reactor environment is not well understood. This study characterizes the room temperature mechanical and electrical properties of a neutron-irradiated CCNZ alloy and compares them to a neutron-irradiated ITER reference heat sink CCZ alloy. Tensile specimens were neutron irradiated in the High Flux Isotope Reactor (HFIR) to 5 dpa between 250 °C and 325 °C. Post-irradiation characterization included electrical resistivity measurements, hardness, and tensile tests. Microstructural evaluation used scanning electron microscopy, energy dispersive x-ray spectroscopy, and atom probe tomography to characterize the irradiation-produced changes in the microstructure and investigate the mechanistic processes leading to post-irradiation properties. Transmutation calculations were validated with composition measurements from atom probe data and used to calculate contributions to the increased electrical resistivity measured after irradiation. Comparisons with CCZ alloys in the same irradiation heat found that the post-irradiated CCNZ and CCZ alloys had comparable electrical resistivity. Although CCNZ alloys suffered more irradiation hardening than CCZ, the overall tensile behavior deviated very little from non-irradiated values in the temperature range studied.

36 MATERIALS SCIENCE↗

AGR-5/6/7 Irradiation Experiment Test Plan

This document presents the test plan for the AGR-5/6/7 irradiation experiment, the combined fifth, sixth, and seventh irradiations of the United States Department of Energy Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The combined irradiation tests will consist of fuel qualification tests (AGR-5/6) and a margin test (AGR-7) for the industrially produced uranium oxycarbide (UCO) tristructural isotropic (TRISO) coated particle fuel developed by the AGR Program. The objective of the AGR-5/6/7 irradiation experiment is to establish the performance of UCO TRISO fuel fabricated using engineering-scale equipment in a production environment under and beyond normal operating conditions. Specifically, the goals of the AGR 5/6/7 irradiation experiment are: 1. Irradiate reference design fuel containing low-enriched UCO TRISO fuel particles to support fuel qualification. 2. Establish the operating margins for the fuel beyond normal operating conditions. 3. Provide irradiated fuel performance data and irradiated fuel samples for post-irradiation examination and safety testing. 4. Support the refinement of fuel performance and fission product transport models with online, post-irradiation examination and safety test data. In order to achieve the test objectives, the AGR-5/6/7 experiment will be irradiated in the northeast flux trap position of the Advanced Test Reactor at Idaho National Laboratory. The larger diameter of the northeast flux trap provides greater flexibility for test train design, significantly enhancing the capability for the combined irradiations. The test train contains five separate and independently controlled and monitored capsules. The middle capsule contains the AGR 7 margin test fuel. The two bottom and two top capsules contain the AGR 5/6 fuel. Each capsule contains a variable number of fuel compacts filled with UCO TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Determining the Effects of Neutron Irradiation on the Structural Integrity of Additively Manufactured Heat Exchangers for Very Small Modular Reactor Applications, DOE Final Report (Project # 19-16980)

Auburn University (AU) teamed with the University of Missouri Research Reactor (MURR) and Kansas State University (KSU) to determine how to best use laser-powder bed fusion (L-PBF) additive manufacturing (AM) methods for generating radiation resistant nickel-based superalloys, Inconel alloy 625 and 718, for special purpose reactor (SPR) or very small modular reactor (vSMR) heat-exchangers (HeXs). Compact, conformal, and durable HeXs that are tolerant of extreme environments are needed for supporting the technical maturity of next-generation, portable compact reactors. AM is an enabler for realizing this new wave of HeXs – providing a means to make customizable hot and cold stream architectures with novel flow path geometries (e.g., tortuous channels with non-uniform, asymmetric cross-sections) and reduced layer-to-layer contact resistance (i.e., no separate bonding procedure required). AM further enables a more time/cost efficient means for fabricating SPRs by reducing the number of suppliers required for HeX assembly and allowing for on-site HeX fabrication. The project aim has been to better understand how neutron irradiation affects the microstructure and properties of additively manufactured nickel-based superalloys, to accelerate their safe, reliable use in the modular reactor industry. The major objective was to qualify/quantify the microstructure and microhardness of nickel-based superalloys (including Inconel 718 and 625) additively manufactured via the L-PBF process in the neutron-dosed (irradiated) and non-irradiated states over a course of 3 years. Effects of build orientation during L-PBF and post-AM heat treatments on neutron resistance, microstructure and mechanical properties were also investigated. Neutron damage mechanisms via hardening were characterized. This project combined subject-matter experts in AM, mechanical/microstructure metallic part characterization, and neutron irradiation, as well as unique assets and capabilities at AU and MURR at MU, to ensure project results translated to effectively addressing known gaps in nuclear science and engineering. Parts were fabricated using L-PBF systems readily available at AU. Specimens were then irradiated using MURR facilities; a manipulator equipped hot cell was also used to measure material hardness after dosing. MURR, a 10 MW, light-water nuclear reactor, is the largest, highest powered, highest-flux university owned research reactor in the U.S. The major findings in this project provide evidence that AM can serve as an alternative way to build structural components for future advanced small modular reactors using advanced materials like Inconel 625 (IN625) or Inconel 718 (IN718). After full spectrum neutron irradiation, vertically as-built AM IN625 samples were observed to display better resistance towards radiation-induced-hardening defects relative to traditionally machined metals. A Vickers microhardness tester, using settings of 1 kgf and dwell time of 15 seconds per indentation, was used to measure hardness in this study. The as-built, vertically printed samples experienced 1.2% of radiation hardening vs. 5.25% radiation hardening observed in wrought IN625. Another set of IN625 and IN718 samples were exposed to fast neutron irradiation. It was observed that IN718 showed more resistance towards radiation hardening compared to IN625 samples indicating IN718 had a better performance. Results showed that the IN718 samples experienced less change (-2.5 to 3.24 %) in microhardness. On the other hand, IN625 samples underwent more (0.9 to 7.21%) change in microhardness after fast neutron irradiation. AM IN625 samples were irradiated using an ion (proton) beam in cyclotron. The mechanical properties of AM samples post irradiation were compared with wrought samples. The irradiated region on the samples were tested using nano-hardness indention. It was observed that the beam current and time used in this study generated an annealing effect and thus reduced the hardness of the samples. The sum of the project results provide precious insight into how one may minimize radiation hardening in AM materials while maintaining material property constraints. Results should assist engineers in selecting an appropriate heat treatment for AM nickel-based superalloys for increased radiation resistance. Results should increase confidence levels for adopting AM for building nuclear reactor components which perform the same or better than conventionally manufactured components. Fast neutron irradiation testing provided an accelerated means of obtaining radiation effects without making materials radioactive and difficult to handle.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Accelerated Ion Irradiation for Emulation of In-reactor Radiation Damage

Reactor core materials in both fast reactors and LWRs granted life extension must withstand irradiation to high doses at high temperature. Ferritic-martensitic (F-M) alloys are attractive candidates for structural components and fuel cladding in fast and thermal reactors, and high chromium and high nickel-containing austenitic steels are potential replacement alloys for LWR core materials. To reach high doses, self-ion irradiation is conducted with simultaneous He injection into both F-M and austenitic alloys, accompanied by reactor irradiations conducted in the BOR-60 fast reactor to assess the capability of ion irradiation to emulate the evolution of microstructure and mechanical properties in reactor. Computational models for defect cluster evolution are being developed and benchmarked against experimental data to ultimately provide predictive capability for the response of both microstructure (loops, voids, precipitates, etc.), and mechanical properties (hardening, ductility, slip behavior) to irradiation. A research program is aimed at understanding how to emulate neutron irradiation using ion irradiation. Results to date have been focused on microstructure and properties of ion irradiated alloys T91, 800H and Fe21Ce32Ni. Data from single ion irradiations are thus far in good agreement with that expected based on previous ion and neutron irradiation data. Dual ion irradiations and neutron irradiated samples are being evaluated and these will be the focus of future work. (authors)

36 MATERIALS SCIENCE↗