Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Gamma irradiation”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 613 records · Page 34

In situ observation of irradiation-induced enhancement to the desorption pressure of zirconium hydride in a nuclear reactor

We quantify the effect of a nuclear-reactor environment on the hydrogen isotope equilibrium vapor pressure over pure zirconium and zirconium hydride. A vacuum-sealed capsule containing a zirconium foil with 6 atom% deuterium was irradiated at a neutron flux of ~10 14 cm -2 s -1 at the University of Missouri Research Reactor (MURR). The internal stainless-steel (SS) sample holder acted as the heat source via gamma absorption. To measure low desorption pressures in a high-flux environment, we developed a method to transduce pressure from the measured sample temperature during irradiation, calibrating with known deuterium pressures in unirradiated capsules at various heating powers using an internal filament-heated system designed to mimic irradiation-induced heating. Our temperature-pressure transduction method operates similarly to a Pirani or thermocouple pressure gauge. The in-reactor measurements revealed a roughly 4-fold enhancement in desorption pressure after only 6 h of irradiation (~2 × 10 18 cm -2 neutron fluence) compared to thermal desorption in control experiments, indicating a nonthermal contribution from neutron irradiation. The slower temperature/pressure stabilization rate in the reactor suggests that desorption pressure enhancement increases with neutron fluence. Further, this enhancement signifies increased solubility of hydrogen isotopes in zirconium during irradiation. We propose that high-energy neutron collisions with hydrogen isotopes in hydrides lead to their decomposition at lower temperatures, supersaturating the surrounding αZr lattice and resulting in higher desorption pressure, which continues to rise as more hydrides dissolve with increasing neutron fluence.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Induction of genomic instability in TK6 human lymphoblasts exposed to 137Cs gamma radiation: comparison to the induction by exposure to accelerated 56Fe particles

The induction of genomic instability in TK6 human lymphoblasts by exposure to (137)Cs gamma radiation was investigated by measuring the frequency and characteristics of unstable clones isolated approximately 36 generations after exposure. Clones surviving irradiation and control clones were analyzed for 17 characteristics including chromosomal aberrations, growth defects, alterations in response to a second irradiation, and mutant frequencies at the thymidine kinase and Na(+)/K(+) ATPase loci. Putative unstable clones were defined as those that exhibited a significant alteration in one or more characteristics compared to the controls. The frequency and characteristics of the unstable clones were compared in clones exposed to (137)Cs gamma rays or (56)Fe particles. The majority of the unstable clones isolated after exposure to either gamma rays or (56)Fe particles exhibited chromosomal instability. Alterations in growth characteristics, radiation response and mutant frequencies occurred much less often than cytogenetic alterations in these unstable clones. The frequency and complexity of the unstable clones were greater after exposure to (56)Fe particles than to gamma rays. Unstable clones that survived 36 generations after exposure to gamma rays exhibited increases in the incidence of dicentric chromosomes but not of chromatid breaks, whereas unstable clones that survived 36 generations after exposure to (56)Fe particles exhibited increases in both chromatid and chromosome aberrations.

Non-NASA Center↗

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↗

Preliminary Evaluation of Commercial Off the Shelf (COTS) Packing Materials for Flight Medication Dispenser (FMD) Technology Development

With the advent of longer duration space missions, pharmaceutical use in space has increased. During the first 33 space shuttle missions, crew members took more than 500 individual doses of 31 different medications . Anecdotal reports from crew members described medications as generally "well tolerated" and "effective". However, reported use of increased medication doses and discrepancies in ground vs. flight efficacy may result from reduced potency or altered bioavailability due to changes in chemical and/or physical parameters of pharmaceutical stability. Based on preliminary results from a ground-based irradiation and an inflight study on pharmaceutical stability, three susceptible medications, Amoxicillin/Clavulanate and Sulfamethoxazole/trimethoprim antibiotics tablets and promethazine (PMZ), an antihistamine were selected for testing using two types of Oliver-Tolas bags, TPC-1475(Clear) and TPF-0599B (Foil) for radiation Shielding effectiveness. The material composition of the bags included aluminum coated Mylar sheathing coated with multifunctional nanocomposities based on polyethylene with dispersed boron-rich nanophases. Two bags of each medication were irradiated for different time intervals with 14.6 rad/min to achieve 0.1 Gy, 1 Gy and 10 Gy of cumulative radiation dose. Active pharmaceutical content (API) in each medication was determined and results analyzed. No significant difference in API content was observed between control and irradiated samples for both antibiotic tablets suggesting both types of bags may offer protection against gamma radiation; results with PMZ were inconclusive. These preliminary results suggest that Oliver-Tolas TPL-1475 and TPF-0599B materials may possess characteristics suitable for protection against ionizing radiation and can be considered for designing and further testing of FMD technology.

Du, Brian↗

Identifying Challenges in Safeguards for Metallic Fuel Fabrication Facilities

As new advanced reactors gain popularity, there is an increasing interest in metallic fuel fabrication for fast reactors. While metallic fuels themselves are not a new idea, as many of the first reactors employed metallic fuels, new designs, compositions, and fabrication methods are appearing throughout the nuclear community. As the interest grows and facilities are constructed, both domestic and international safeguards will need to be heavily involved to support safeguards-by-design (SBD) measures from the start. This work compiles a review of historical and modern fuel types and fabrication methods, fabrication processes, safeguards gaps, and potential safeguards solutions. Metallic nuclear fuel types have been around for many decades and were included in some of the first reactors including the Experimental Breeder Reactor (EBR)-I and -II, the Fermi 1 reactor, the Integral Fast Reactor (IFR), and the Dounreay Fast Reactor (DFR). These reactors used various compositions including pure uranium (U) metal, U-zirconium (Zr) alloys, plutonium (Pu)-aluminum (Al) alloys, U-fissium (Fs) alloys, U-Pu-Zr alloys, and U-molybdenum (Mo) alloys [1, 2, 3, 4, 5]. These small alloying additions are included to improve the material properties of the pure U metal. The alpha-phase U (stable below 661C) suffers elongation in one direction causing grain boundary cracking and increasing creep rate due to irradiation growth, thermal cycling, and preferential crystal orientation. It is ideal to utilize the gamma-phase U (typically stable above 769C) by adding small amounts of alloying elements such as Zr or Mo to stabilize this phase down to room temperature [3]. Additionally, some research has been focused on U with transuranic (TRU) elements present, typically coming from the used fuel recycling process. Including these elements in fast reactor fuel can aid in the reduction of nuclear waste by burning minor long-lived actinides. However, the additions of TRU elements can cause concerns to arise when trying to fabrication or safeguard metallic fuels. A typical metallic fuel element is shown in Figure 1. Sodium is added into the cladding to create a thermal bond between the fuel slug and cladding wall. The fuel slug is then inserted and the end plug is welded on to the top of the fuel element. A gas plenum is left to create a headspace for gaseous fission products to escape rather than continue to build in the fuel itself [1, 5]. Other fuel element geometries exist as well, such as the Lightbridge twisted cruciform geometry shown in Figure 2 [6]. This design allows for better cooling performance and provides room for fuel rod swelling without impacting the fuel rod diameter. There are many different fabrication methods for metallic fuels, which is one of the many benefits of these fuel types. Many of these fabrication methods are relatively easy and cost-efficient. The most popular fabrication method is injection casting, sometimes called vacuum induction melting (VIM), shown in Figure 3 [4, 8, 9, 7, 10]. This method was largely used for EBR-II fuel fabrication. The injection casting system is contained inside of a vessel consisting of a Y2O3-coated graphite crucible surrounded by an induction coil with ZrO2-coated quartz molds suspended above the crucible. The fuel feedstock is placed inside of the graphite crucible and melted using the induction furnace. The induction furnace utilizes a dual frequency with the high frequency melting the feedstock and the low frequency causing stirring of the melted feedstock to form a homogeneous mixture. The mixture is heated to approximately 1600C in an argon environment. The vessel is evacuated and then the quartz molds are lowered into the graphite crucible containing the molten metal and the vessel is repressurized to inject the metal fuel upwards into the molds. The molds are removed and then shattered to release the fuel slugs. This fabrication method was used to fabricate 39,000 metallic fuel pins for EBR-II. While injection casting has been the most common metallic fuel fabrication method throughout the decades, many other methods have been explored including low-pressure gravity casting, microwave casting, continuous casting, centrifugal casting, coextrusion, and many others [11, 12, 8, 13, 14, 15]. Some of these methods aim to mitigate challenges that arise with americium (Am) volatilization during the casting process for TRU-containing fuel feedstocks, an issue with injection casting. Coextrusion is one of the methods explored at the Idaho National Laboratory (INL) and has been utilized for the initial fabrication tests of Lightbridge's unique fuels, as well as other metallic fuels with cladding coextruded. In this process, large billets are formed and machined and then inserted into a molten salt bath for approximately 30 minutes. The billets are then loaded into the extrusion press and extruded. This process can be seen in Figure 4 [15].

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neutron Dosimetry for the University of Central Florida (UCF3) Irradiation in ATR

PNNL project 74242 involves the analysis of neutron fluence monitors irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00236287 and Statement of Work No. 17370, Rev. 0, PNNL Analysis of NSUF Flux and Melt Wire Capsules. This report is for the University of Central Florida (UCF)-3 third stage experiment which was conducted in position B8 of the ATR. Three other irradiations included in the scope of work are reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the UCF-3 assemblies at INL 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, and 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 8 fluence monitor locations. 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↗

Neutron Dosimetry for the Colorado School of Mines (CSM 16-10584) Irradiation in ATR

PNNL project 74242 involves the analysis of neutron fluence monitors irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00236287 and Statement of Work No. 17370, Rev. 0, PNNL Analysis of NSUF Flux and Melt Wire Capsules. This report is for the Colorado School of Mines (CSM 16-10584) experiment which was conducted in position B5 of the ATR. Three other irradiations included in the scope of work are reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the CSM assemblies at INL 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, and 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 12 fluence monitor locations. 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↗

A portable and monoenergetic 24 keV neutron source based on 124Sb-9Be photoneutrons and an iron filter

A portable monoenergetic 24 keV neutron source based on the 124Sb-9Be photoneutron reaction and an iron filter has been constructed and characterized. The coincidence of the neutron energy from SbBe and the low interaction cross-section with iron (mean free path up to 29 cm) makes pure iron specially suited to shield against gamma rays from 124Sb decays while letting through the neutrons. To increase the 124Sb activity and thus the neutron flux, a >1 GBq 124Sb source was produced by irradiating a natural Sb metal pellet with a high flux of thermal neutrons in a nuclear reactor. The design of the source shielding structure makes for easy transportation and deployment. A hydrogen gas proportional counter is used to characterize the neutrons emitted by the source and a NaI detector is used for gamma background characterization. At the exit opening of the neutron beam, the characterization determined the neutron flux in the energy range 20–25 keV to be 6.00±0.30 neutrons per cm2 per second and the total gamma flux to be 245±8 gammas per cm2 per second (numbers scaled to 1 GBq activity of the 124Sb source). A liquid scintillator detector is demonstrated to be sensitive to neutrons with incident kinetic energies from 8 to 17 keV, so it can be paired with the source as a backing detector for neutron scattering calibration experiments. This photoneutron source provides a good tool for in-situ low energy nuclear recoil calibration for dark matter experiments and coherent elastic neutrino-nucleus scattering experiments.

Dark Matter detectors (WIMPs, axions, etc.)↗

Neutron Dosimetry for the Boise State University JW 15-8242 Irradiation in ATR (Rev.1)

PNNL project 79550 provides for the analysis of neutron fluence monitors irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory in accordance with MPO 00269673and Statement of Work No. 19704, Rev. 0, PNNL Analysis of NSUF Flux Capsules. This report is for the Boise State University (BSU) JW 15-8242 irradiations which were conducted in positions A6, A7 and A8 of the ATR. We previously reported data for this experiment after about 1 dpa in a report issued in January 2021 (1). This report is for longer irradiations to about 3 dpa. Other experiments included in the statement of work for this project will be reported separately. The neutron fluence monitors were prepared by PNNL and loaded into the BSU 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 or x-ray analysis, and 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↗

Early and Late Damages in Chromosome 3 of Human Lymphocytes After Radiation Exposure

Tumor formation in humans or animals is a multi-step process. An early stage of cancer development is believed to be genomic instability (GI) which accelerates the mutation rate in the descendants of the cells surviving radiation exposure. GI is defined as elevated or persistent genetic damages occurring many generations after the cells are exposed. While early studies have demonstrated radiation-induced GI in several cell types as detected in endpoints such as mutation, apoptosis and damages in chromosomes, the dependence of GI on the quality of radiation remains uncertain. To investigate GI in human lymphocytes induced by both low- and high-LET radiation, we initially exposed white blood cells collected from healthy subjects to gamma rays in vitro, and cultured the cells for multiple generations. Chromosome aberrations were analyzed in cells collected at first mitosis post irradiation and at several intervals during the culture period. Among a number of biological endpoints planned for the project, the multi-color banding fluorescent in situ hybridization (mBAND) allows identification of inversions that were expected to be stable. We present here early and late chromosome aberrations detected with mBAND in chromosome 3 after gamma exposure. Comparison of chromosome damages in between human lymphocytes and human epithelial cells is also discussed

Sunagawa, Mayumi↗

Effect of Additives: Gamma-Ray Sensors Based on Ionizing Organic Nanocomposites

We have developed ionizing organic-based composites which have demonstrated great promise for radiation sensing. Nickel oxide has been proven as an active material for detecting high energy radiation. The oxidation state of unusual oxides such as nickel oxide in nanocomposites of ionizing organics changes much faster than thin film or bulk, and hence increases the sensitivity for radiation sensing. The resistivity of the oxide composite increases following sequential irradiation processes because of the decrease in holes' concentration. In this paper, we will present the effect of additional oxidizing agent on the morphology, processing and sensing of gamma-rays by oxides - urea based nanocomposites. It was observed that addition of MnO2 decreases the sensitivity. However, it recovers to original properties after irradiation much faster than undoped composites.

Vishall Dayal↗

Assessment of Self-interrogation Safeguards Signatures for Pebble Bed Reactor Fuel

New reactor designs, such as pebble-bed reactors, present challenges for the safeguarding of fissile material. Due to the multi-pass fuel circulation design and the non-uniform path a pebble may traverse through a pebble-bed reactor, there will be variations in the irradiation history of spent fuel pebbles. The standard approach of estimating actinide quantities based on fission gamma spectra or neutron emission paired with depletion modeling may yield uncertainties too large for safeguards and material accountancy. This project investigated, through modeling and experimentation, the potential of neutron self-interrogation of spent fuel pebbles as an innovative method to implement materials accountability. As an example, our feasibility studies indicate that the mass of U-235, U-238, Pu-239 and Pu-241 can be predicted to 4.1, 0.86, 13 and 13 % accuracy, respectively, when measuring 100 closely packed end-of-life spent fuel pebbles over approximately 12 days using a 4π counting geometry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The nucleosynthesis of deterium and helium-3

A new model of the creation of D and 3He in supernova of the first generation was considered. It is based on the idea that a supernova event leads simultaneously to acceleration of particles in the shock wave front and to their interactions with supernova atmosphere matter. The D, 3He, Li, Be, B and bypassed isotopes are created in these interactions. The cosmic abundances of the bypassed isotopes with A or = 113 allowed to determine the integral proton flux I sub p (E sub p 25 MeV) = 5 x 10 to the 21/sq cm the spectral index - gamma = 3. The calculations of the D and 3He yields in various nuclear reactions at these irradiation conditions show that only the 4He(p,d)3He reaction leads to cosmic abundances of these isotopes on the assumption that all matter has been exposed.

Lavrukhina, A. K.↗

Computation of Domain-Averaged Shortwave Irradiance by a One-Dimensional Algorithm Incorporating Correlations between Optical Thickness and Direct Incident Radiation

A one-dimensional radiative transfer algorithm that accounts for correlations between the optical thickness and the incident direct solar radiation is developed to compute the domain-averaged shortwave irradiance profile. It divides the direct irradiance into four components and treats the direct irradiance in two separate, clear and cloudy columns to account for the fact that clouds attenuate the direct irradiance more than clear-sky. The horizontal inhomogeneity of clouds in the cloudy column is treated by the gamma weighted two-stream approximation, which assumes that the optical thickness of clouds follows a gamma distribution. The algorithm inputs the cloud fraction, cumulative cloud fraction as a function of height, and a parameter expressing the shape of the probability density function of the cloud optical thickness distribution in addition to inputs required for a two-stream radiative transfer model. These cloud property inputs can be obtained using ground- and satellite-based instruments. Therefore, the algorithm can treat realistic cloud overlap features and horizontal inhomogeneity of clouds in a framework of one- dimensional radiative transfer. Heating rates computed by the algorithm using cloud fields generated by cloud resolving models agree with those computed with a Monte Carlo model. If optical properties in computational layers that divide a vertically extensive cloud are correlated, the irradiance profile computed by the algorithm further improves.

Kato, S.↗

Remote Radiation Sensing Using Aerial and Ground Platforms

Remote sensing of ionizing radiation has a significant role in waste management, nuclear material management and nonproliferation, and radiation safety. Robotic platforms can surpass the number of tasks that are achieved by humans. With this technique, the operator's radiation exposure can be decreased. Remote sensing allows for the evaluation and monitoring of radiological contamination. Gamma-ray and neutron sensors were integrated onto the robotic platforms. This approach allows for the radiation sensor data to be dynamically tracked and mapped thus enabling further analysis of the radiation flux in temporal and spatial domains. The goal is to complete scheduled tasks while the robot is being irradiated. To achieve this, electronic components must be shielded and radiation hardened. CZT Detector: Cadmium Zinc Telluride (CZT) detector technology has been a promising solution for gamma-ray and x-ray measurements. Detector data is transferred to the Odroid minicomputer that controls and powers the module via the USB. Robot Operating System (ROS) was utilized for data acquisition and data fusion. The Mariscotti method was employed for the spectrum analysis. A function was programmed in ROS for the automatic identification of photopeaks. CLYC Detector: A Cs{sub 2}LiYCl{sub 6}:Ce{sup 3+} (CLYC) detector was used for simultaneous medium-resolution gamma-ray measurements and neutron counting. A 2.54 cm diameter photomultiplier tube (PMT) was equipped with a high voltage supply and a miniature digitizer. Gamma-ray excitation: fast core-to-valence luminescence (CVL) with 1 ns decay constant, and prompt Ce{sup 3+} emission with 50 ns decay constant. Neutron excitation: slow cerium self-trapped excitation (Ce{sup 3+} STE), 1000 ns decay constant. Radiation Source Localization: Maximum Likelihood Estimation (MLE) and gradient-based methods were used to locate the position of a radiation source based on measured radiation intensities. Multi-Particle Transport Code FLUKA: Estimation of radiation damage of the electronic components is important in order to optimize the robot's operational time while it is irradiated. Displacement per atom (DPA) represents the radiation damage in materials exposed to the ionizing radiation. Various shielding layers of different thickness t were analyzed (< 5% statistical error). The model of the controller of the UAS was designed in FLUKA. Conclusion: CZT and CLYC detectors were integrated onto the robotic platforms. Radiation source localization and contour mapping using robotic platforms were studied. Functions for data analysis and fusion were developed in ROS. FLUKA code was utilized to analyze DPA values. Layers of low-density and high-density materials were used to shield the UAS electronics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

LLNL measurements of thermally irradiated HEU sample and saltwater-matrixed HEU sample

Two scintillation vials, kindly provided by the PNNL team, were received in the LLNL radiochemistry building on June 10, 2024; one containing approximately 2 mL dissolved thermally-irradiated HEU, and the other containing dried salt from irradiated seawater. A quick, semiquantitative screening count of the salt showed that the activation product activities were very low, with the highest activity being ~1 Bq 24 Na. The dissolved HEU solution was transferred to a Prindle vial (LLNL standard counting geometry) and weighed. Approximately 2 mL 3 M HNO 3 was added so that the solution completely covered the bottom of the vial and Gamanal software would be able to accurately generate an efficiency curve. This sample, called “PNNLFP24” is the irradiated HEU solution received, gravimetrically diluted by a factor of 1.7356 ± 0.0003, and was sent for quick gamma counting prior to further modification. Meanwhile the salt was quantitatively transferred to a 250 mL polyethylene bottle and dissolved in 160 mL 3 M HNO 3 . After counting, 50 µL PNNLFP24 was aliquoted for Resonance Ionization Mass Spectrometry (RIMS) analysis, and then 10.0 mL of the salt solution (6.03% of the total salt, gravimetrically) was added to the remaining PNNLFP24 solution to make “PNNLFPSW24”. After weighing, two 100 µL aliquots were removed for RIMS and DES, and 3 mL was aliquoted for microfluidic chemistry. The remaining “PNNLFPSW24” solution was weighed and proceeded for singles and coincidence counting on the MCBOS system.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO2), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Post-Irradiation Examination on MiniFuel UCO and UO 2 TRISO Particles Irradiated in HFIR at High Power

Post-irradiation examination (PIE) of MiniFuel compacts was conducted at Oak Ridge National Laboratory (ORNL) under the Nuclear Science User Facilities project in collaboration with Kairos Power (KP) to evaluate the performance of tristructural-isotropic (TRISO) particles under high particle power and fluoride-salt-cooled high-temperature reactor (FHR)-relevant conditions. MiniFuel compacts containing low-enriched uranium oxide-uranium carbide (LEUCO), low-enriched uranium dioxide (LEUO 2 ), and natural UCO (NUCO) kernels were irradiated for four cycles at ORNL’s High Flux Isotope Reactor (HFIR) at target temperatures between 500°C and 900°C. Post irradiation, the experiment was disassembled at ORNL to recover the MiniFuel subcapsules, which were subsequently punctured to measure fission gas release. Subcapsule disassembly allowed the recovery of components of interest, such as silicon carbide (SiC) thermometry, fuel specimens, fission product sinks, and SiC spacers. The experimental irradiation temperature was confirmed by analyzing the SiC thermometry via dilatometry. PIE on the fuel specimens included gamma spectrometry and deconsolidation leach burn leach, which were complemented by imaging techniques such as x-ray computed tomography, optical microscopy, and electron microscopy. The PIE results provide insight into TRISO particle integrity, fission product retention, coating performance, and kernel migration, informing fuel qualification for application in KP’s FHR concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗