Results for “Irradiation testing” — Page 39

NASA NTRS1987

Spectrometric test of a linear array sensor

A spectroradiometer which measures spectral reflectivities and irradiance in discrete spectral channels was tested to determine the accuracy of its wavelength calibration. This sensor is a primary tool in the remote sensing investigations conducted on biomass at NASA's Goddard Space Flight Center. Measurements have been collected on crop and forest plants both in the laboratory and field with this radiometer to develop crop identification and plant stress remote sensing techniques. Wavelength calibration is essential for use in referencing the study measurements with those of other investigations and satellite remote sensor data sets. This calibration determines a wavelength vs channel address conversion which was found to have an RMS deviation of approximately half a channel, or 1.5 nm in the range from 360 to 1050 nm. A comparison of these results with those of another test showed an average difference of approximately 4 nm, sufficiently accurate for most investigative work.

Brown, Kenneth S.↗

DOE OSTI2023

PARFUME/BISON Fission Product Release Predictions versus AGR-3/4 Heating Test Measurements

The fuel performance modeling codes PARFUME (PARticle FUel ModEl) and BISON were used to predict the release of fission products silver, cesium, and strontium from as-irradiated fuel compacts containing tristructural isotropic (TRISO) coated particles during heating tests post irradiation. The AGR-3/4 fuel compacts were irradiated as part of the third and fourth series of planned experiments to support the Advanced Gas Reactor (AGR) program. The heating tests were conducted at temperatures between 1200°C and 1700°C to simulate reactor accident conditions. The measured fission product release fractions from the heating tests were compared to modeling predictions calculated by PARFUME and BISON to evaluate how the codes compare to experimental results. Comparisons between the experimental measured fission product release fractions from silver, cesium and strontium indicate that both modeling codes overpredict the fission product release fractions demonstrating that the diffusivities used in the codes are overestimated. This results in a conservative estimate predicted by the codes when evaluating the fission product release relative to experimental data as it pertains to silver, cesium, and strontium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Destructive PIE and Safety Testing of Six AGR-5/6/7 Capsule 2 Compacts

This study evaluates fission product retention and particle failure mechanisms in AGR-5/6/7 Capsule 2 uranium carbide and uranium oxide (UCO) tristructural isotropic (TRISO) fuel under high-temperature gas reactor accident-relevant conditions using high-temperature safety tests and destructive postirradiation examination. Three Capsule 2 compacts were held isothermally at 1600°C for approximately 300 hours and one compact at 1800°C for approximately 300 hours; two additional compacts were examined in the as-irradiated state. Post-test deconsolidation–leach–burn–leach (DLBL) quantified nuclide inventories in matrix and particles. Individual particles were surveyed for radioisotope inventories, and microanalytical approaches resolved microstructural evolution and fission product distributions within the coating layers. At 1600°C, no krypton was detected above the minimum detectable limit, and cesium releases were far below a single particle equivalent, indicating the absence of full TRISO failure or SiC failures. Silver releases were limited and primarily reflected depleted postirradiation inventories, consistent with prior compact-level exams indicating substantial in-pile 110mAg loss. At 1800°C, cumulative 134Cs release of approximately 2.5 particle equivalents and delayed 85Kr totaling approximately 0.53 particle equivalents were consistent with one full TRISO failure and two SiC failures. Europium and strontium releases were roughly one order of magnitude higher than at 1600°C and comparable to AGR-1/AGR-2 high-temperature tests, with sustained late-hold rates indicating diffusion through intact coatings coupled with matrix depletion. Overall, AGR-5/6/7 Capsule 2 UCO fuel demonstrated fission product retention during safety testing consistent with prior AGR campaigns, while distinctive in-pile 110mAg depletion and measurable 1600°C europium loss motivate targeted follow-on studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Thermal-hydraulic and Fuel Performance Scoping Studies of a Flowing Water Capsule in TREAT

The restart of the Transient Reactor Test (TREAT) facility has provided a much needed capability for integral transient testing of nuclear fuel. This testing is necessary to qualify new fuel concepts such as those developed under the Accident Tolerant Fuel (ATF) program, increasing the burnup limit of light water reactor (LWR) fuels, or a variety of other programs under the Advanced Fuels Campaign (AFC). The ATF campaign has been the main driver behind the development and implementation of a variety of capsules for TREAT experiments. The Separate-Effect Test Holder (SETH) was a dry capsule that enabled testing of ATF concepts under reactivity-initiated accident (RIA) heating conditions. Following SETH, the Static Environment Rodlet Transient Test Apparatus (SERTTA) was developed to enable RIA testing in a static water environment. In efforts to support the need for future Loss-of-Coolant Accident (LOCA) tests, the Transient Water Irradiation System for TREAT (TWIST) capsule has been developed that allows for water to drain from around the fuel rod and lower the pressure to simulate LOCA conditions. All these capsules that have been developed for LWR fuel testing all lack the capability for forced convective cooling which in some applications may limit their ability to test under prototypic conditions. A design and modeling effort has been started to modify the TWIST capsule by adding a flow tube and impeller that can create flowing coolant conditions for the fuel rod. RELAP5-3D and BISON models are being used to study the differences between RIA, LOCA, Anticipated Operational Occurrences (AOO), LWR power cycling, and other transient scenarios under flowing conditions capable in the flowing capsule and the current stagnant water capsules (SERTTA and TWIST). The scoping study will provide guidance on the needed capabilities for the flowing capsule and the limitations of the currently developed capsule for LWR testing in TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

The effects of high energy deuteron ion beam irradiation on the tensile behavior of HT-9

Ion beam implantations are widely performed to understand the effects of irradiation-induced displacement damage on nuclear structural materials. However, the volume of material that can be investigated is often limited by the maximum energies of accelerator facilitates, leading to a limitation in the thickness of samples whose mechanical properties can be evaluated. The Lawrence Berkeley National Laboratory’s 88-Inch Cyclotron offers a wide range of ions and energies, allowing for material ion implantations at larger scales than typical. Four HT-9 SS-J-geometry tensile specimens were polished and then irradiated with deuterons at the 88-Inch Cyclotron to doses of approximately 0.2 dpa prior to small scale tensile testing. Here, the results from this study show irradiation hardening characterized by the tensile test results and black dot irradiation defects. Additionally, a comprehensive look at low temperature irradiations of high-Cr F/M steels is presented and our results show agreement with the available data.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Neutron Dosimetry for the GE Hitachi 16-10393 Irradiation in ATR

PNNL project 74242 involves the analysis of neutron fluence monitors and melt wires irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00236287 and Statement of Work (SOW) No. 17370, Rev. 0, PNNL Analysis of NSUF Flux and Melt Wire Capsules. This report is for the GE Hitachi 16-10393 irradiation which was conducted in positions B11 of the ATR. Three other irradiations included in the SOW will be reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the GE Hitachi assemblies prior to irradiation. Following irradiation, the capsules were returned to PNNL for analysis. The neutron dosimetry capsules were opened, the flux wires were removed for gamma analysis. The measured activities were used to determine the activation rates for various activation products. Following suitable corrections, the measured activation rates were used to adjust calculated neutron spectra at each fluence monitor location. The adjusted neutron spectra were then used to determine displacement per atom (dpa) and gas production for irradiated materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

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↗

NASA NTRS1987

Fracture surfaces of irradiated composites

Electron microscopy was used to analyze the fracture surfaces of T300/934 graphite/epoxy unidirectional off-axis tensile coupons which were subjected to 1.0-MeV electron radiation at a rate of 50 Mrad/h for a total dose of 10 Grad. Fracture surfaces from irradiated and nonirradiated specimens tested at 116 K, room temperature, and 394 K were analyzed to assess the influence of radiation and temperature on the mode of failure and variations in constituent material as a function of environmental exposure. Micrographs of fracture surfaces indicate that irradiated specimens are more brittle than nonirradiated specimens at low temperatures. However, at elevated temperatures the irradiated specimens exhibit significantly more plasticity than nonirradiated specimens.

Milkovich, Scott M.↗

DOE OSTI2020

Fluence Capsules for the SAM-2 Experiment MPO# 00239221/SOW-17494, Rev. 0

The statement of work (SOW) applies to the fabrication and analysis of flux wire capsules that will be used in the SAM-2 experiment. This experiment is a drop-in design slated for irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The 8 neutron fluence capsules will be used to determine the neutron exposure that the specimens have received. For Task 1 Pacific Northwest National Laboratory (PNNL) will prepare eight (8) neutron fluence monitors for irradiation in the ATR. Task 2 will involve analysis port-irradiation; the monitors will be returned to PNNL for analysis in FY21 after irradiation. The analysis report will include the neutron fluences and radiation damage calculations for all irradiated positions in the experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Effects of 5-Ion Beam Irradiation and Hindlimb Unloading on Metabolic Pathways in Plasma and Brain of Behaviorally Tested WAG/Rij Rats

A limitation of simulated space radiation studies is that radiation exposure is not the only environmental challenge astronauts face during missions. Therefore, we characterized behavioral and cognitive performance of male WAG/Rij rats 3 months after sham-irradiation or total body irradiation with a simplified 5-ion mixed beam exposure in the absence or presence of simulated weightlessness using hindlimb unloading (HU) alone. Six months following behavioral and cognitive testing or 9 months following sham-irradiation or total body irradiation, plasma and brain tissues (hippocampus and cortex) were processed to determine whether the behavioral and cognitive effects were associated with long-term alterations in metabolic pathways in plasma and brain. Sham HU, but not irradiated HU, rats were impaired in spatial habituation learning. Rats irradiated with 1.5 Gy showed increased depressive-like behaviors. This was seen in the absence but not presence of HU. Thus, HU has differential effects in sham-irradiated and irradiated animals and specific behavioral measures are associated with plasma levels of distinct metabolites 6 months later. The combined effects of HU and radiation on metabolic pathways in plasma and brain illustrate the complex interaction of environmental stressors and highlights the importance of assessing these interactions.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗

NASA NTRS1984

The Solar Constant, Climate, and Some Tests of the Storage Hypothesis

Activity related modulation of the solar constant can have practical consequences for climate only if storage is involved, as opposed to a detailed balance between sunspot blocking and facular reemission. Four empirical tests are considered that might distinguish between these opposing interpretations: monochromatic measurements of positive and negative flux; comparison of modelled and measured irradiance variations; the interpretation of secular trends in irradiance data; and the direct test of an anticipated signal in climate records of surface air temperature. The yet unanswered question of the role of faculae as possible reemitters of blocked radiation precludes a definitive answer, although other tests suggest their role to be minor, and that storage and an 11 year modulation is implicated. A crucial test is the behavior of the secular trend in irradiance in the declining years of the present activity cycle.

Eddy, J. A.↗

NASA NTRS2019

Analyses and Methods of Solid Rocket Motor Material Irradiation at Marshall Space Flight Center

The search for life on other worlds is among humanity’s greatest endeavors. Europa represents the most probable location to discover extraterrestrial life in our solar system, owing to its surface composition of ice covering a liquid water ocean, warmed by the tidal forces of its orbit around Jupiter. Unfortunately, the Jovian system hosts the most intense planetary radiation environment in the solar system due to the charged particles, namely electrons and protons, trapped by Jupiter’s immense magnetic field. Any mission that attempts to approach or land on Europa must survive this radiation environment [Hand et al, 2017]. Radiation effects were identified as a priority risk to the successful development of a de-orbit stage and solid rocket motor (SRM) early in the Europa Lander De-orbit Stage project concept. The effects of primary concern tend to occur very near the outer surface of the SRM. The charged particles deposit their energy quickly and are mostly stopped in the outer metallic case, but a significant portion of radiation penetrates through the bondline and outer propellant regions. High doses of ionizing radiation are known to cause significant changes to mechanical properties of many materials, especially polymers. For polymers such as the rubber-like materials (elastomers) in a solid rocket motor, the primary damage mechanism is known as cross-linking, in which ionization causes the restructuring of the matrix of long polymer chains. Ionization energy breaks the long polymer chains and allows formation of new cross-linked bonding sites. This hardens and often strengthens the polymer, but at the cost of decreased flexibility (or modulus). Propellant, insulation, liner, and pyrotechnic materials were identified as higher risk items, and so were irradiated at Marshall Space Flight Center (MSFC) for investigation of changes in mechanical and ballistic properties. This process required significant levels of analysis to evaluate how the radiation environment evolves within the spacecraft during the mission, and also to evaluate how dose is delivered into test articles within the irradiating facilities.

Caffrey, Jarvis↗

NASA NTRS2019

Analyses and Methods of Solid Rocket Motor Material Irradiation at Marshall Space Flight Center

The search for life on other worlds is among humanity’s greatest endeavors. Europa represents the most probable location to discover extraterrestrial life in our solar system, owing to its surface composition of ice covering a liquid water ocean, warmed by the tidal forces of its orbit around Jupiter. Unfortunately, the Jovian system hosts the most intense planetary radiation environment in the solar system due to the charged particles, namely electrons and protons, trapped by Jupiter’s immense magnetic field. Any mission that attempts to approach or land on Europa must survive this radiation environment [Hand et al, 2017]. Radiation effects were identified as a priority risk to the successful development of a de-orbit stage and solid rocket motor (SRM) early in the Europa Lander De-orbit Stage project concept. The effects of primary concern tend to occur very near the outer surface of the SRM. The charged particles deposit their energy quickly and are mostly stopped in the outer metallic case, but a significant portion of radiation penetrates through the bondline and outer propellant regions. High doses of ionizing radiation are known to cause significant changes to mechanical properties of many materials, especially polymers. For polymers such as the rubber-like materials (elastomers) in a solid rocket motor, the primary damage mechanism is known as cross-linking, in which ionization causes the restructuring of the matrix of long polymer chains. Ionization energy breaks the long polymer chains and allows formation of new cross-linked bonding sites. This hardens and often strengthens the polymer, but at the cost of decreased flexibility (or modulus). Propellant, insulation, liner, and pyrotechnic materials were identified as higher risk items, and so were irradiated at Marshall Space Flight Center (MSFC) for investigation of changes in mechanical and ballistic properties. This process required significant levels of analysis to evaluate how the radiation environment evolves within the spacecraft during the mission, and also to evaluate how dose is delivered into test articles within the irradiating facilities.

Caffrey, Jarvis↗

DOE OSTI2020

Evaluation of the NSUF Reactor Activation and Damage (RAD) Calculator damage component

The Nuclear Science User Facilities (NSUF) provide researchers access to nuclear research facilities and expertise all over the United States and at the Belgian Center for Nuclear Research (SCK/CEN). One special capability of the NSUF is providing access to nuclear research and test reactors often used for conducting material testing experiments. These services are provided at no cost to the researcher through a competitive, peer-reviewed process. With so many proposals being submitted simultaneously, an NSUF information management resource called the Combined Materials Experiment Toolkit (CoMET) was implemented to speed up the review process and give users access to the information and tools necessary for designing meaningful experiments and writing the best proposals possible. One of CoMET’s tools is the Reactor Activation and Damage (RAD) Calculator. For those with little or no background in it, neutron irradiation in research and test reactors can be a challenging subject to include in a proposal. That is why the RAD Calculator was created to assist users with providing estimates of irradiation and post-irradiation conditions, enabling them to select the best neutron irradiation and post-irradiation examination facilities for reaching their experiment goals. The RAD Calculator is a general scoping tool that estimates radiation damage in terms of DPA, in addition to the resulting radioactivity of the material post-irradiation. This report evaluates the neutron damage component of the RAD Calculator. The evaluation of the activation component of the calculator is covered in report INL/EXT-20-58080.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

PM-HIP Mechanical Data Archive

This is the mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). Data in this archive include uniaxial tensile testing data, fractography of selected tensile bars, and nanoindentation. The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300ºC and 400ºC through the DOE Office of Nuclear Energy, Nuclear Science User Facilities (NSUF) project 15-8242.

Wharry, Janelle↗

DOE OSTI2026

Fractography

This is the mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). Data in this archive include uniaxial tensile testing data, fractography of selected tensile bars, and nanoindentation. The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300ºC and 400ºC through the DOE Office of Nuclear Energy, Nuclear Science User Facilities (NSUF) project 15-8242.

Wharry, Janelle↗

DOE OSTI2026

G91-P_400C_1DPA_(ID-403)

This is the mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). Data in this archive include uniaxial tensile testing data, fractography of selected tensile bars, and nanoindentation. The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300ºC and 400ºC through the DOE Office of Nuclear Energy, Nuclear Science User Facilities (NSUF) project 15-8242.

Wharry, Janelle↗

DOE OSTI2026

G91-P_400C_3DPA_(ID-670)

This is the mechanical testing data archive from a neutron irradiation campaign of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP). Data in this archive include uniaxial tensile testing data, fractography of selected tensile bars, and nanoindentation. The irradiation campaign was designed to facilitate a direct comparison of PM-HIP to conventional casting or forging. Five common nuclear structural alloys were included in the campaign: 316L stainless steel, SA508 pressure vessel steel, Grade 91 ferritic steel, and Ni-base alloys 625 and 690. Irradiations were carried out in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to target doses of 1 and 3 displacements per atom (dpa) at target temperatures of 300ºC and 400ºC through the DOE Office of Nuclear Energy, Nuclear Science User Facilities (NSUF) project 15-8242.

Wharry, Janelle↗