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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Radiation Stability of MOF Engineered Particles

Pacific Northwest National Laboratory (PNNL), in collaboration with Flibe Energy, demonstrated the fabrication of metal organic frameworks (MOFs) into engineered beads to manage off-gases released from advanced reactors, including molten salt reactors. Based on the commercial availability and easy to synthesis at scale, PNNL down-selected two MOF sorbents known as CuBTC and UiO-67. The sorbents were synthesized, characterized, and fabricated into mechanically robust engineered beads (composites) using wet granulation method using polymer as a binder. Further, MOF powders and and mechanically robust composites were exposed to gamma radiation using Co-60 source and characterized irradiated MOFs to demonstrate the structural stability. Among the two MOF sorbents identified, the CuBTC MOF was shown to be very promising for noble gas management at low temperature (193 K) compared to UiO-67 MOF. The high adsorption capacity of the synthesized composites was correlated to the high surface area and pore volume of the CuBTC MOF. Both MOFs retain structural integrity and long-range order even after exposure to 1000 kGy of radiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Specifications of Legacy U(Pu)Zr Metallography Data

The DOE Nuclear Energy Advanced Reactor Technologies (ART) Program has supported the creation of several databases with information describing the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data is one of the most important types of PIE data being collected, organized and stored in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three main sets of metallography data. The first set is the metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins measured in the Hot Fuel Examination Facility (HFEF); the second set is the metallography data from EBR-II irradiated fuel pins measured in the Alpha-Gamma Hot Cell Facility (AGHCF); the third set is the metallography data from transient tested fuel pins (the transients tests include the out-of-pile tests and TREAT tests) measured in AGHCF. Since both the second and third sets of data were measured in AGHCF, they are governed by the same specification. The metallography data in the databases are digital images scanned from either positive or negative photos. The quality of the images, including their resolution and contrast, relies on the preserved quality of the pictures and the scanning conditions. To analyze the microstructure of a fuel pin, a series of preparatory steps must be undertaken. These include sectioning, epoxy mounting, mechanical grinding and polishing, and often etching. The resulting samples were then transferred to a secondary hot cell, or glovebox (depending the strength of radiation field) for microscopy examination. The specifications provided herein focus on the metallography examinations. The sample grinding, polishing and etching processes are also discussed. The procedures involving sectioning and epoxy mounting are out of the scope of the current specification; details can be found in the corresponding operation manuals. If more data are collected and added to the ART Fast Reactor Databases, this specification will be updated to accommodate the additional data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Determination of Scale Bar for AGHCF Metallography Data

The DOE Nuclear Energy Advanced Reactor Technologies (ART) Fast Reactor Program (FRP) has supported the development of several databases containing information on the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data represents one of the most critical types of post-irradiation examination (PIE) data being collected, organized, and archived in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Metallic Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three principal sets of metallography data. The first set comprises metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins examined in the Hot Fuel Examination Facility (HFEF). The second set consists of metallography data from EBR-II irradiated fuel pins examined in the Alpha-Gamma Hot Cell Facility (AGHCF). The third set includes metallography data from transient-tested fuel pins (including both out-of-pile furnace tests and TREAT tests) examined in AGHCF. Since both the second and third sets were generated in AGHCF, they are governed by identical specifications. The metallography data in the databases consist of digital images scanned from either positive or negative photographic films. To analyze the microstructure of a fuel pin, a series of preparatory steps are required, including sectioning, epoxy mounting, mechanical grinding and polishing, and etching. Following sample preparation, specimens are transferred for metallographic examination. The AGHCF and HFEF metallography data were generated using optical microscopes manufactured by Leitz and Bausch and Lomb (B&L). Images were recorded on Polaroid film at preset magnifications. Magnification verification for the Leitz and B&L metallographs was conducted every two months prior to 1989 and at least every six months from 1989 through the conclusion of the IFR program. Magnifications determined from imaging of microslide standards were compared to the instrument settings for magnifications ranging from 50× to 500×. If the magnifications determined from standards deviated from the instrument settings, adjustments were made to the bellows extension until agreement was achieved. The specifications for AGHCF and HFEF legacy metallography data have been established based on available hard-copy and digital records, most of which have been incorporated into the data repositories associated with FIPD, OPTD, and TREXR. Detailed specifications including hard-copy records, digitized records, cutting diagrams and sectioning schemes, high-magnification photographs, photomosaics (composites), information tags, scale bars, and magnification verification procedures can be found in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Fission Product Quantification and Burnup Analysis via Gamma Spectrometry of the AGR-5/6/7 TRISO Fuel Experiment

Five capsules containing varying configurations of TRISO fuel compacts were irradiated in the most recent Advanced Gas Reactor (AGR) TRISO test campaign, AGR-5/6/7. As part of the post-irradiation examination (PIE) of this work all the fuel compacts and capsule graphite holders were measured with the Precision Gamma Scanner (PGS) to determine isotopic activities and locations using gamma spectrometry. Measuring the compacts allowed for the comparison of physics model depletion calculations with the measured compact activity to validate models. These measurements also are used to find compacts of interest with low fission product retention that may be tested in subsequent PIE. Measurements of the holders can confirm fission product distribution and migration. Of the isotopes measured, Cs-137 and Cs-134 were used to estimate location of failed particles as well as determine a measured burnup estimation. Results from Capsule 1 compacts and the holder indicate many fuel particle failures occurred. There are high amounts of cesium in the holder near Stack 9 of Capsule 1 that corresponds to low fractions?measured-to-calculated (M/C) ratio?of cesium remaining in those compacts. All other capsules had cesium retention ratios near 1. The isotope 110mAg had less consistent results. Capsule 2 compacts exhibited a M/C of 0.20, much lower than anticipated for the temperatures these compacts underwent. Capsule 3 experienced much higher temperatures, >1300C and had an average M/C of 0.75. These are the opposite results than what was predicted by past AGR experiments for the relationship between temperature and isotope retention. Capsule 4 and 5 showed good fission product retention and agreement with the physics model predictions. The burnup results from the PGS measurements showed strong agreement between the physics models and the measured values. Using a ratio of Cs-134/Cs-137 to calculate the burnup match closely with the physics models. Additionally, the Cs-137 burnup results were less accurate but extremely close to the model results.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Leveraging Machine Learning Capabilities for the Characterization of Irradiated Uranium: A Case Study of Analysis Methods for Nuclear Safeguards and Nuclear Forensics

Nondestructively determining the initial enrichment of irradiated uranium is a complex and laborious multivariable problem due to the presence of fission products. This work demonstrates the capabilities of machine learning to analyze gamma-ray spectral data to determine initial enrichment without knowledge of the decay time of the sample. The approach developed is agnostic to the particular scenario and is applicable to a wide variety of applications in nuclear forensics and nuclear safeguards. We irradiated 5 mg uranium standard reference materials at discrete enrichment values ranging from 0.02% to 97% 235 U (weight percent) in UT Austin’s Nuclear Engineering Teaching Laboratory TRIGA Mark II 1.1 MW research reactor, allowed each to decay for 8 hours, and then measured each sample via gamma-ray spectrometry for 50 hours post-irradiation yielding 1,400 individual gamma-ray spectra discretized into 8,192 energy bins. We then trained decision trees models to analyze individual gamma-ray spectra and estimate the associated initial enrichment without knowledge of the time since end of irradiation. We evaluated the performance of the models with a reserved test set not used for training or calibrating the model. A decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 2.3% (weight percent 235 U). Next, we implemented a principal component analysis pre-processing routine of the gamma-ray spectrometry data to reduce the dimensionality of the dataset from 8,192 channels in the spectrum to 10 principal components while retaining over 99% of the inherent variance in the data. Decision tree models constructed with these data demonstrated decreased mean absolute error in enrichment determination, reduced computational time, and decreased complexity. A single decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 0.05% (weight percent 235 U). Furthermore, we analyzed these models with learning curves to ensure that overfitting did not occur. The capabilities provided by these models can be naturally extended to other application-focused measurements in the fields of nuclear safeguards, nuclear forensics, and nuclear non-proliferation.

Drescher, Adam↗
DOE OSTI2025

A methodology for corrosion testing ODS steels in liquid tin under reactor conditions in HFIR

Liquid metal–based divertor concepts have promising attributes for application in fusion reactors. Liquid tin is being considered as one of those liquid metals due to its excellent thermophysical properties. However, tin is extremely corrosive to steels at elevated temperatures, and the coupled effects of corrosion with neutron irradiation for application in fusion reactors have not been quantified. Researchers at Oak Ridge National Laboratory (ORNL) have designed and irradiated a series of capsules in the High Flux Isotope Reactor (HFIR) to help address gaps in the literature with respect to coupled corrosion and neutron irradiation effects. The capsules are filled with solid tin designed to melt due to the gamma heating in HFIR and allow interaction with the specimens. The capsules contain a SiC thermometer for post-irradiation temperature verification. Five capsules were irradiated in HFIR for 10.5 days, accumulating 0.93 dpa in the FeCrAl specimens. Post-irradiation temperature verification was performed on the SiC thermometers, and temperatures were observed to be higher than what was expected from thermal models. An explanation for the model underprediction is given herein. In conclusion, the capsules described within this article are the first to irradiate liquid metal within the HFIR flux trap and demonstrated the safety basis for continued liquid coolant studies in HFIR.

FeCrAl↗
DOE OSTI2022

Fast neutron activation of ubiquitous materials

Nuclear explosions expose ubiquitous materials to large numbers of neutrons, producing a variety of radioactive isotopes. To simulate such phenomena from both fission and thermonuclear explosions, we irradiated 29 different targets with approximately 3 and 14 MeV neutrons and measured the beta-delayed gamma rays using germanium detectors. Therefore, for each neutron energy, the expected radioisotopes, half-lives, and gamma ray energies were deduced. From measurements of the ratios of activities of the radionuclides produced by neutron irradiations, we were able to identify several materials that are particularly sensitive to the neutron energy spectra.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Godiva-IV Dosimetry Exercise 2022 Preliminary Results

Integral Experiment Request (IER) 538 is part of a series of dose characterization and nuclear accident dosimetry (NAD) exercises performed under the Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). This is the second NAD exercise using the Godiva-IV critical assembly and the third NAD exercise overall. The participating laboratories provided their own dosimeters that were mounted on the Lawrence Livermore National Laboratory (LLNL) BOttle Manikin ABsorption (BOMAB) phantoms and aluminum plates. The BOMABs and plates were placed at two, three, and four meters away from the center of Godiva. Alongside the NADs, there was a LLNL Passive Neutron Spectrometer (PNS), Atomic Weapons Establishment (AWE) PNS, and Y-12 Sphere present to measure the neutron dose from Godiva. Two irradiations were conducted to test the NAD performance from each laboratory and assesses their performance to the DOE-STD-1098-2017 part 515 criteria. Neutron and gamma doses were measured prior to this exercise. This work presents a model for the neutron and gamma dose respectively to serve as the reference value. A code written in C/C++/ROOT was used to fit the measured neutron and gamma dose with the new models. It was assumed that the neutron and gamma doses are proportional to the change in temperature of Godiva after a burst irradiation. Uncertainties for the reference values were calculated using error propagation of the model’s parameters. Preliminary results (within twenty-four hours) and final results were compared for each laboratory. On average of all the participating laboratories, 32% of neutron doses and 78% of gamma doses were outside the DOE standards. One laboratory did not report their dose readings and were not included in this average. There is a bias for a lower neutron dose and a higher gamma dose based on the distribution of results. In comparison with the past Godiva-IV NAD exercise, there is an improvement in neutron dose readings by 20%.

BOttle Manikin ABsorption (BOMAB)↗
DOE OSTI2023

Radiation tolerance studies of the HV-mux GaNFETs for the HL-LHC ATLAS ITk Strip detector

For the High-Luminosity upgrade of the LHC, the current ATLAS inner detector will be replaced with a new silicon charged-particle tracker, the ITk, which consists of the ITk Pixel and the ITk Strip subdetector. The high voltage multiplexing (HV-Mux) GaNFETs are radiation-tolerant transistors that permit switching off high voltage to malfunctioning sensors on the ITk Strip modules. To ensure the reliability of the GaNFETs in the high radiation environment expected at the HL-LHC, a sample of the production batch was exposed to gamma radiation. Finally, the GaNFETs were characterized pre-irradiation and post-irradiation, and monitored during irradiation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Efficiency Calibration of FGMS Cold Traps

This plan describes tests to determine the efficiency calibrations for the Fission Gas Monitoring System (FGMS) by the measurements of 85Kr and 133Xe gamma rays. FGMS collects fission product gases released from irradiated Advanced Gas Reactor tristructural-isotropic-coated particle fuel specimens during post-irradiation heating in the Fuel Accident Condition Simulator (FACS) furnace. FGMS is composed of two charcoal-filled, liquid-nitrogen cooled traps that are each monitored by a high purity germanium (HPGe) detector. All irradiated Advanced Gas Reactor fuel specimens contain 85Kr (t1/2 = 10.8 year); however, significant short-lived fission products such as 133Xe (t1/2 = 5.2 day) and 131I (t1/2 = 8.0 day) have decayed away by the time they reach the FACS for testing. In order to measure the releases of these short-lived fission products during safety testing in the FACS furnace, a re-irradiation of samples is necessary in the Neutron Radiography Reactor. This requires the FGMS be calibrated for the fission gas 133Xe, in addition to 85Kr. Because any re-irradiated fuel specimen will have both 85Kr and 133Xe present, the collection efficiency from a mixture of krypton and xenon needs to be determined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Radiobiological Studies Using Cs-137 Gamma or 320 kV X Rays

There are numerous self-shielded research irradiators used in various facilities throughout the United States. The irradiators employ radioactive sources containing either 137 Cs or 60 Co and the irradiators are used for a variety of radiobiological investigations involving cellular and/or animal models. A report from the National Academy of Sciences described security issues associated with particular radiation sources and the desire for their replacement with suitable X-ray irradiators. One possible replacement would be a 320 kV X-ray irradiator. The participants in this research successfully performed in vivo radiobiological studies involving mice exposed to filtered (HVL ≈ 4 mm Cu) 320 kV X rays. Two publications (one published and one submitted at the publishing of this report) documenting key findings are provided in Appendices A and B of this report. The 320 kV X rays were found suitable for in vivo (in mice) cell survival studies and are expected to be suitable for bone marrow transplantation studies using mice but this needs to be experimentally validated.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

Dense Pair and Gamma-Ray Creation Using Ultra-Intense Lasers (Final Technical Report)

Lasers with intensity >1.4x10 18 W.cm -2 irradiating solid targets couple 10-50% of its energy to “hot electrons” near the critical surface, with temperature kT > mc 2 . When these hot electrons impact high-Z target ions (e.g. Au, Pt), they emit copious bremsstrahlung gamma-rays and create e + e - pairs via the Trident process. For targets thicker than ~0.1mm, the gamma-rays create secondary pairs via the Bethe-Heitler processes. The concept of using lasers to create pairs was first studied by Shearer et al (1973). Motivated by the rapid advance of short-pulse ultra-intense lasers based on chirped-pulse amplification, Liang (1994) first proposed irradiating solid Au targets with laser intensity ≥10 20 W.cm -2 to create dense e + e - pairs. Using particle-in-cell (PIC, Birdsall and Langdon, 1991) simulations, Liang et al (1998, 2002) estimated that high-energy PW lasers can in principle achieve ultra-high in-situ pair densities. Subsequent studies supported this idea and found that for gold foils thicker than ~50 μm, the pair yield is dominated by the Bethe-Heitler (BH, Heitler 1954) process. Cowan et al (1999, 2000) using the LLNL Nova PW-laser to irradiate 125μm gold foils first observed pair creation, followed by Chen et al (2009) who demonstrated copious BH pair creation using the Titan, Omega-EP and Orion lasers. However, the emerging e + /e - ratio of these experiments was ≤ few %, and the pair density was only 10 13 /cc so that the pair jet transverse size R(~mm) was < pair skin depth c/ω + (=8πn+e 2 /m) 1/2 ). This was insufficient to qualify as a bona fide ”pair plasma”.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Performance testing of a Compton suppressed coincidence measurements using the Advanced Radionuclide Gamma-spectrOmeter

A Compton-suppressed multidimensional gamma-spectrometer is being developed within the Shallow Underground Laboratory (SUL) at Pacific Northwest National Laboratory (PNNL, USA). This Advanced Radionuclide Gamma-spectrOmeter (ARGO) system consists of two Broad Energy Germanium (BEGe) detectors, surrounded by a 12-detector sodium iodide (NaI) Compton suppression shield, situated inside a low-background lead and copper shield, fitted with a cosmic veto background reduction system. The detector has advanced functionality, including operation in single or combined detector mode, coincidence mode and Compton suppressed coincidence mode. Validation experiments measuring a neutron irradiated natural uranium sample, have demonstrated order of magnitude improvements in gamma detection sensitivity compared to conventional gamma-spectrometry techniques. Importantly this research demonstrates the feasibility of processing and analysing high-multiplicity datasets and the potential for high-sensitivity radionuclide measurements.

ARGO↗
DOE OSTI2022

Post-irradiation characterization of a high burnup mixed oxide fuel rod with minor actinides

The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). Here, it was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Irradiation Testing of Ultrasonic Transducers

Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of numerous parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 (E> 0.1 MeV). This test will be an instrumented lead test; and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. By characterizing magnetostrictive and piezoelectric transducer survivability during irradiation, test results will enable the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers

Daw, J.↗
DOE OSTI2025

Impact of corrosion layer composition and thermal pretreatment on the radiolytic generation of molecular hydrogen from aluminum alloys

To fully evaluate the feasibility of extending the dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters, the amount of radiation-induced molecular hydrogen (H 2 ) generated and the impact of absorbed radiation dose on the ASNF corrosion layer composition must be accurately assessed. Here, in this study, we report a slowing in the rate of radiolytic H 2 generation for pre-corroded AA1100 and AA6061 aluminum alloy coupons irradiated to up to 53 MGy of absorbed cobalt-60 gamma dose. By exploring a variety of thermal pretreatment conditions for AA6061 coupons, we find that the “steady-state” yield of H 2 depends more on the aluminum alloy used than on the treatments applied prior to dry storage. Scanning electron microscopy and positron annihilation lifetime spectroscopy techniques also provided evidence for gamma radiation-induced defects in the corrosion layers of the investigated aluminum alloy coupons for high absorbed doses (~50 MGy), the consequences of which on cladding integrity and H 2 generation should be explored in future works.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2025

Radiation Characterization Summary: Godiva IV Critical Assembly Environments at the In-Core, Top Hat, 1m, and 2m Irradiation Locations

This document presents the facility-recommended characterization of the neutron, prompt gamma ray, and delayed gamma ray radiation fields at the Godiva IV critical assembly at the National Criticality Experiments Research Center (NCERC). The environments assessed include the In-Core location, a location on the Top Hat, 1m away from the assembly, and 2m away from the assembly. The neutron, prompt gamma ray, and delayed gamma ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma ray fluence profiles on the Top Hat surrounding the critical assembly. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulse operations are presented with conversion examples.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Improvements to the 115g/115m Cd Nuclear Data

Repeated measurements of the isotope 115m Cd by Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL), and Atomic Weapons Establishment (AWE) have revealed a consistent discrepancy between beta and gamma counting. This suggests that a significant disparity exists between the true value of the isotope’s gamma-ray branching ratios and the values reported by the National Nuclear Data Center (NNDC). Enriched 114 Cd was irradiated using thermal neutrons to produce a high-purity 115m/g Cd source for counting and analysis using traditional singles gamma-ray spectroscopy, liquid scintillation counting, gas-proportional counting, and analysis using the new Gamma-Alpha-Beta-Gamma (GABγ) and Gamma-Alpha-Beta Radio-Isotope EvaLuator (GABRIEL) coincidence detection systems. Using the D-T fusion generated neutrons, reduces the wait time required to let the short-lived 115Cd ground state decay away by leveraging the 1:1 production ratio of the ground and metastable state of 115 Cd from the 115 In (n,p) reaction relative to the 10:1 production ratio of the ground and metastable isotopes of 115 Cd by thermal neutron capture on 114 Cd. However, the much higher flux in from the Washington State University TRIGA reactor is a more rapid method of production of 115m/gCd but requires access to enriched 114 Cd. High purity 115m/g Cd samples were analyzed on regular intervals over a period of 140-days. The activity of 115m Cd measured by beta and gamma show a clear discrepancy using the current best known gamma-ray decay branching ratios. Based on these observations and measurements, a set of new branching ratio recommendations have been produced using conventional counting techniques and the advanced GAB? and GABRIEL instruments.

07 ISOTOPE AND RADIATION SOURCES↗