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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.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Effect of Surface Temperature on the Distribution and Reactivity of Rh Active Sites for CO Oxidation

Single–atom catalysts are a fast–emerging area in which late–transition metal atoms are supported on oxides, metals, and carbonaceous supports. They show great promise for selective chemical reactions due to their well–defined active sites, and significantly reduce precious metal loading. However, oxide–supported single–atom catalysts have drawbacks such as deactivation due to sintering, and there are on–going debates about the exact nature of the active sites. Herein we report a model system where the surface temperature during Rh deposition on a copper oxide support dictates the Rh size distribution. Here our temperature–programmed desorption experiments reveal that Rh clusters perform CO oxidation at lower temperature than single atoms, and isotopic labelling demonstrates differences in the mechanism in that the C–O bond is broken and reformed on clusters, as opposed to a pure Mars van Krevelen mechanism for CO oxidation at the single–atom Rh sites. We further examine the cluster size distributions with scanning tunneling microscopy and X–ray photoelectron spectroscopy which confirm the presence of both clusters and single atoms for the lower temperature deposition and only single atoms for the higher temperature deposition. Together, these results provide a fundamental understanding of the differences in the CO oxidation mechanism on Rh atoms and clusters.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutron radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Baseline Postirradiation Examination of the FUTURIX-FTA Experiments

Postirradiation Examination (PIE) results from the baseline examination of the FUTURIX-FTA irradiation experiment are presented in this report. The FUTURIX-FTA irradiation was designed to study the fuel performance of candidate transmutation fuels. The FUTURIX-FTA experiment was conducted in the true fast neutron spectrum conditions of the Phénix sodium fast test reactor in France. FUTURIX-FTA contained several compositions that were also irradiated as part of the AFC-1 test in cadmium filtered positions in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) that approximate the neutron spectrum of a fast reactor. This experiment was a joint collaboration between the Department of Energy (DOE) in the U.S. and the Commissariat á l'Energie Atomique et aux Energies Alternatives (CEA) in France. A joint agreement between DOE and CEA was established in 2004 to conduct an irradiation experiment of metallic and nitride fuels in the Phénix reactor at CEA. Both the FUTURIX-FTA tests and the AFC-1 tests were conducted for the DOE Advanced Fuels Campaign (AFC) that is part of Fuel Cycle Research & Development (FCRD) program. These programs seek to develop and demonstrate the technologies needed to transmute long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology. The long term goal of this work is to reduce the radiotoxicity of future high level waste repositories. Post-irradiation examination of irradiated fuels experiments provides data related to in reactor fuel performance and input into future fuel design choices. The compositions of the FUTURIX-FTA pins were designed to test if minor actinides could be incorporated into the fuel of a fast reactor without significantly changing fuel performance observed for previously well studied fuels. The fuel performance of metallic fuel in fast reactors has been well documented. Likewise the irradiation performance of nitride fuels in fast reactors has also been documented. However the existing database for nitride fuels is much smaller than the database for metal fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Improving Zr cross sections: 90 Zr(n, y) [Slides]

This presentation covers key take-aways such as the first of 4 isotopes (90,91,92,94) to be measured. 91 Zr is currently being measured at GELINA. Once all measurements are done a join evaluation with SAMMY will be undertaken. Further, there will be production of new evaluated resonance parameters and covariances. New data will be used to improve evaluation information, benefitting transport calculations of reactors, waste storage, etc.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Validating Mixtures of 233 U, 235 U, and 239 Pu for the Sum-of-Fractions Method

The Sum-of-Fractions method is a technique used to assure that homogeneous mixtures of fissile and fissionable isotopes are below a minimum margin of k eff or reactivity. Current work by Pacific Northwest National Laboratory examines different mixtures of 233 U, 235 U, and 239 Pu to determine critical mass limits for mixtures of transuranic actinides lacking a validation basis. To provide a validation basis for these limits, the work presented here describes the results of a sensitivity and uncertainty analysis of various mixtures of these isotopes in various concentrations moderated and reflected by light water and polyethylene. The TSUNAMI-1D sequence in the SCALE code system was used to generate sensitivity coefficients for three different concentrations of mixtures of 233 U, 235 U, and 239 Pu. The TSUNAMI-IP sequence was then used for similarity assessment (c k ) with critical benchmark experiment sensitivity data files (SDFs) from the Oak Ridge National Laboratory Verified, Archived Library of Inputs and Data and the Nuclear Energy Agency SDF database. The VADER sequence in SCALE was used for statistical testing and to generate upper subcritical limits from the data to develop a basis for validating critical mass limits.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Atmospheric CO 2 estimates for the Miocene to Pleistocene based on foraminiferal $δ^{11}$B at Ocean Drilling Program Sites 806 and 807 in the Western Equatorial Pacific

Constraints on the evolution of atmospheric CO 2 levels throughout Earth's history are foundational to our understanding of past variations in climate. Despite considerable effort, records vary in their temporal and spatial coverage and estimates of past CO 2 levels do not always converge, and therefore new records and proxies are valuable. Here we reconstruct atmospheric CO 2 values across major climate transitions over the past 16 million years using the boron isotopic composition (δ 11 B) of planktic foraminifera from 89 samples obtained from two sites in the West Pacific Warm Pool, Ocean Drilling Program (ODP) Sites 806 and 807, measured using high-precision multi-collector inductively coupled plasma mass spectrometry. We compare our results to published data from ODP Site 872, also in the Western Equatorial Pacific, that goes back to 22 million years ago. These sites are in a region that today is near equilibrium with the atmosphere and are thought to have been in equilibrium with the atmosphere for the interval studied. We show that δ 11 B data from this region are consistent with other boron-based studies. The data show evidence for elevated pCO 2 during the Middle Miocene and Early to Middle Pliocene, and reductions in pCO 2 of ~200 ppm during the Middle Miocene Climate Transition, ~250 ppm during Pliocene Glacial Intensification and ~50 ppm during the Mid-Pleistocene Climate Transition. During the Mid-Pleistocene Transition there is a minimum pCO 2 at marine isotopic stage (MIS) 30. Our results are consistent with a coupling between pCO 2 , temperature and ice sheet expansion from the Miocene to the late Quaternary.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Fully convolutional neural network for removing background in noisy images of uranium bearing particles

A fully convolutional neural network (FCN) was developed to supersede automatic or manual thresholding algorithms used for tabulating SIMS particle search data. The FCN was designed to perform a binary classification of pixels in each image belonging to a particle or not, thereby effectively removing background signal without manually or automatically determining an intensity threshold. Using 8,000 images from 28 different particle screening analyses, the FCN was trained to accurately predict pixels belonging to a particle with near 99% accuracy. Background eliminated images were then segmented using a watershed technique in order to determine isotopic ratios of particles. A comparison of the isotopic distributions of an independent data set segmented using the neural network, compared to a commercially available automated particle measurement (APM) program developed by CAMECA, highlighted the necessity for effective background removal to ensure that resulting particle identification is not only accurate, but preserves valuable signal that could be lost due to improper segmentation. The FCN approach improves the robustness of current state-of-the-art particle searching algorithms by reducing user input biases, resulting in an improved absolute signal per particle and decreased uncertainty of the determined isotope ratios.

Tarolli, Jay G. (ORCID:0000000307369057)↗
DOE OSTI2026

Mark-18A Target Material Recovery Program: Initial Hot Startup

The Savannah River National Laboratory (SRNL) has been tasked by the National Nuclear Security Administration to recover highly valued isotopes from irradiated Mark-18A (Mk-18A) targets. The Savannah River Site (SRS) has sixty-five Mk-18A targets available for the recovery of the high valued materials. The Mk-18A targets were manufactured with Pu-242 then irradiated under high neutron flux in K-Reactor at the SRS from 1968 to 1978. The sixty-five Mk-18A targets are currently stored in the SRS L-Area Basin and will be removed one at a time and individually transported to SRNL. Upon receipt at SRNL, the Mk-18A target material will be removed from the confinement, dissolved, chemically separated, and calcined to a stable oxide. The flowsheet is designed to recover the plutonium as well as the trivalent actinides. The original targets were manufactured with varying quantities of Pu-242, ranging from 5.2 grams to 121.5 grams. Taking a graded approach to process start up, the lowest loaded target (FT-80-03) was selected as the first target to be received and processed at SRNL. As operational experience and knowledge is gained from processing targets, higher loaded targets will be selected for processing to increase the quantities of valuable isotopes recovered. Due to dose concerns, the receipt and processing of the Mk-18A targets is performed in the SRNL Shielded Cells Facility. The targets are stored in a double J-can confinement in the L-Area Basin. A specially designed cask was procured for transport of the targets from L-Area to SRNL. Once received at SRNL, the targets are loaded into the back of Cell 7 and resized as they enter the cell. The resized targets (1/4 length) are then processed one at a time through the following processes: caustic dissolution and filtration, acidic dissolution and filtration, elutable Reillex anion exchange, diglycolamide (DGA) resin extraction, and DGA calcination. This processing results in two product streams. The first is an aqueous plutonium solution which is removed from the shielded cells and taken to a glovebox for further purification and conversion to an oxide. The second is a calcined oxide product containing the Am and Cm as well as other lanthanide fission products which is removed from the shielded cells using a bagless transfer system. Both materials are packaged for shipment to Oak Ridge National Laboratory (ORNL). This paper will discuss the operating experience, lessons learned, and results from initial process hot startup.

Armstrong, Christopher [Savannah River National La↗
DOE OSTI2022

Laboratory Directed Research and Development Program Activities (LDRD 2021 Annual Report)

Each year, Brookhaven National Laboratory (BNL) is required to provide a report of its completed Laboratory Directed Research and Development Program (LDRD) projects to the Department of Energy (DOE) Office of Scientific and Technical Information in accordance with DOE Order 413.2C Chg1 (MinChg) dated August 2, 2018. This report provides a detailed look at the scientific and technical activities for each of the LDRD projects funded by BNL in FY 2021, in fulfillment of that requirement. In FY 2021, the BNL LDRD Program funded 68 projects, 24 of which were new starts, at a total cost of $17.1M. The investments that BNL makes in its LDRD program support the Laboratory’s strategic goals. BNL has identified six scientific initiatives that define the Laboratory’s scientific future and that will enable it to realize its overall vision. This requires simultaneous excellence in all aspects of BNL’s work – from science and operations, to external partnerships with the local, state, and national communities, and beyond. This is enabled by safe, efficient, and secure operations; by an unwavering commitment to a diverse, equitable, and inclusive environment, including workforce development, both with staff and reaching out to the community; and by a strong focus on renewed infrastructure. The six scientific initiatives are: (1) Nuclear Physics: The Electron Ion Collider, (2) Clean Energy and Climate, (3) Quantum Information Science and Technology, (4) Discovery Science Driven by the Human-AI-Facility Integration, (5) High Energy Physics: Building for Discovery, and (6) Accelerating Isotope Production: Ensuring the Nation’s Supply is Secure. The funded projects support BNL’s six scientific initiatives and priority programs as well as new areas of research and competencies at the Laboratory that are consistent with the Laboratory’s vision and mission. In total, these LDRD investments supported 80 postdoctoral researchers and graduate students in whole or in part and resulted in 136 publications and 3 awards. This Program Activities Report represents the future of BNL science; it is an impressive body of exploratory work that investigates many scientific and technical directions in support of the DOE and BNL missions.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Comparison of Light-Weight Radioisotope Heater Unit (LWRHU) Historical Testing to Special Form Testing Requirements

This paper compares the Light-Weight Radioisotope Heater Unit (LWRHU) historical testing to Department of Transportation (DOT) Special Form criteria per 49 CFR Part 173, Shippers-General Requirements for Shipments and Packages, Section 469, Tests for special form Class 7 (radioactive) materials as specified in DOE-STD-1027, Hazard Categorization of DOE Nuclear Facilities. Compliance with the aforementioned criteria allows for the exclusion of radioactive sources from the facility’s nuclear material inventory. As part of a new program at Los Alamos National Laboratory (LANL), LWRHUs are planned to be moved into TA-55, PF-5 Power Source Assembly Area (PSAA) for assembly and testing activities. The PSAA Facility is categorized as a low hazard, non-nuclear facility per DOE STD-1027 and is described in TA55-PF-5-FHC-11-001, Facility Hazard Categorization of the PSAA at PF-5, TA-55. The Facility Hazard Categorization document states, there must be documentation that radioactive sources are engineered to pass the special form testing specified by DOT in 49 CFR 173.469, Tests for special form Class 7 (radioactive) materials or ANSI N43.6, Sealed Radioactive Sources - Classification. This paper documents the extensive historical environmental safety testing of the LWRHU and demonstrates its ability to survive potential launch safety accident scenarios. The testing results are used herein to demonstrate through a comparative analysis that the robust design of the LWRHU meets or exceeds the special form testing specifications in 49 CFR 173.469, enabling the LWRHU to be transported and handled in PF-5 in support of testing and assembly program.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

National‐Scale Assessment of Total Gaseous Mercury Isotopes Across the United States

Abstract With the 2011 promulgation of the Mercury and Air Toxics Standards by the U.S. Environmental Protection Agency, and the successful negotiation by the United Nations Environment Program of the Minamata Convention, emissions of mercury (Hg) have declined in the United States. While the declines in atmospheric Hg concentrations in North America are encouraging, linking the declines to changing domestic and global source portfolios remains challenging. To address these research gaps, the U.S. Geological Survey initiated the first national‐scale effort to establish a baseline of total gaseous mercury stable isotope values at 31 sites distributed across the United States. Results indicated that unique Hg sources, such as Hg evasion from an elemental Hg contaminated site or free tropospheric intrusions in high altitude sites, were distinguishable from background atmospheric values. Minor gradients were observed across the nation, with regions of heavy industrial activity demonstrating lower δ 202 Hg, but no consistent changes in other isotopes such as Δ 199 Hg and Δ 200 Hg were observed. Furthermore, δ 202 Hg was impacted by foliar uptake and senescence but trends varied between forested regions in the northeastern and midwestern United States. These data demonstrate regional emission sources and other environmental variables can impact total gaseous Hg (TGM) isotope values, highlighting the need to characterize atmospheric Hg isotopes over larger geographical areas to evaluate changes related to national and international Hg regulations.

Tate, Michael T.↗
DOE OSTI2023

Developing a Facility NMAC Plan

The table presented below suggests the basic information that should be covered in a facility NMAC Plan for an NMAC program that is designed for nuclear security. The topics are appropriate for and should be addressed by all facilities in their NMAC Plans. They are appropriate for NMAC Plans for nuclear power plants, research reactors, fuel manufacturing facilities, facilities that produce medical isotopes, and other facilities. The difference is in the intensity with which the various measures are applied and the thoroughness of the description of the application (i.e., the program requirements). The robustness of a facility NMAC program and the content of its NMAC Plan should be graded in accordance with the type of facility and the category of its nuclear material.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Post-irradiation Examination of Eurofer-97 Steel Irradiated to 20 dpa at 200–400°C in HFIR under the EUROfusion (ORNL-KIT) Collaboration Program

The Oak Ridge National Laboratory/Karlsruhe Institute of Technology (ORNL/KIT) collaboration focuses on research involving irradiation experiments and post-irradiation examinations (PIE) of isotopically modified Eurofer-97 steels for fusion reactor applications. This collaboration capitalizes on ORNL’s expertise in radiation effects on materials and its capabilities in irradiation and PIE. The research aims to qualify Fe-9Cr-based Eurofer-97 steel under simulated fusion conditions, which include high doses and high transmutation rates. A unique isotopic modification technique is utilized in the research, wherein high-transmutation isotopes such as Fe-54 and Ni-58 are alloyed into the Eurofer steel to align with the expected helium production rates in fusion reactor conditions. This report presents the results of post-irradiation mechanical testing activities for the EUROFER-97/2 specimens after irradiation to ~20 dpa at various irradiation temperatures. The irradiation doses and temperatures for the ES capsules ranged from 18.4–20.4 dpa and 202–256°C, respectively. The mechanical property datasets, obtained through baseline testing and PIE of the tensile and fracture specimens, include microhardness data from 93 irradiated and non-irradiated tensile specimens and 28 irradiated bend bar specimens, uniaxial tensile property data from 45 irradiated and non-irradiated tensile specimens, and fracture toughness data from the bend bar specimens. Further data analyses provide statistical information on the microhardness values and tensile properties, as well as the reference ductile-brittle transition temperature (T 0 ) data.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Proceedings for the Workshop on Applied Nuclear Data Activities 2024

The Workshop for Applied Nuclear Data Activities (WANDA) is designed to increase communication among nuclear data (ND) users in multidisciplinary federal programs, ND producers, ND funders, and other ND experts. It also presents an opportunity to cross-pollinate ideas as well as introduce ND gaps identified by federal programs to ND experts and ND capabilities to the various federal ND users. WANDA 2024 included five technical sessions, three of which focused on Fusion Energy Sciences (FES)—FES Fusion Neutronics, FES Tritium Production, and FES Material Damage—and two stand-alone sessions—Isotopes and Targetry for Nuclear Data and Uncertainty Quantification. The FES sessions successfully brought new voices to the WANDA discussions, expanding the application space in which nuclear data are critical. FES programs need accurate nuclear data with realistic uncertainty quantification to properly estimate, for example, shielding, activation, tritium production, helium production, structural material integrity, and superconducting magnet operation. This includes a variety of projectile (neutrons, photons, charged particles) and target atoms. One of the action items common to all the FES sessions was a need to perform sensitivity studies to identify the prioritization of nuclear data needs. The Isotopes and Targetry session highlighted the many capabilities available to produce high-quality targets for nuclear data measurements, including 3D printing with spherical powders, combustion synthesis coupled with spin coating & electrospraying, inkjet printing, and isotopic doping. These new methods open doors for more accurate measurement, but it was also stressed that sample characterization following any method of fabrication is of the highest importance to accurately interpret nuclear data measurement results that used that sample. The Uncertainty Quantification (UQ) session was broken into two categories: nuclear data uncertainty quantification and the use of that uncertainty quantification. Thematic to the UQ session was the loss of information when going from nuclear data measurement, to evaluation, to evaluated file, and finally to neutron transport calculations. Current evaluated ND libraries typically only contain covariances, which assume that the probability distributions are Gaussian. Beyond being a simplified assumption for many evaluations, this can lead to negative values on many observables when attempting to sample the covariance. The covariance format, however, is very efficient in that a simple set of linear equations can transform uncertainty from parameters or cross sections to the application of interest. Focused collaboration is needed between nuclear data evaluators and nuclear data users to ensure that needs are being met.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Understanding the effect of specimen size on the properties of fine-grain isotropic nuclear graphite for irradiation studies: Mechanical properties

Programs at Oak Ridge National Laboratory are investigating the effects of neutron irradiation in the High Flux Isotope Reactor (HFIR) on properties of fine-grain isotropic nuclear graphites. Specimens were irradiated in the HFIR flux trap, which requires using specimens that are smaller than the sizes recommended by ASTM standards. Here, this work has investigated whether subsized specimens can be used to reliably predict strength behavior. The results show that (1) tensile strength was loosely coupled to specimen volume, (2) the compression specimens had a slightly higher strength for larger volumes, (3) the uniaxial flexural strength was nearly independent of specimen effective area for one orientation whereas the other orientation showed a slight strength–size scaling Weibull behavior, and (4) equibiaxial flexural strength followed traditional strength–size scaling Weibull behavior. These results suggest that fine-grain graphite is more similar to ceramics than to medium- and large-grained graphite with respect to both the strength-limiting flaws (Griffith theory) and the effect of machining on the strength behavior. This work shows that using specimens that are smaller than ASTM recommendations are suitable for investigating the strength changes caused by neutron irradiation because the primary interest is the relative change in the strength distribution parameters rather than their absolute values.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Understanding the effect of specimen size on the properties of fine-grain isotropic nuclear graphite for irradiation studies: Physical, electrical, thermal properties

Programs at Oak Ridge National Laboratory are investigating the effects of neutron irradiation in the High Flux Isotope Reactor (HFIR) on properties of fine-grain nuclear graphites. Specimens were irradiated in the HFIR flux trap, which required using specimens that are smaller than the sizes recommended by ASTM standards. Here, this work has investigated whether subsized specimens can be used to reliably determine the measured physical (density, elastic moduli, and electrical resistivity) and thermal properties (coefficient of thermal conductivity and thermal conductivity). The results show the specimen sizes in this investigation had a minimal influence on any of the measured properties. Even for the properties where size influenced the measured value (density and thermal conductivity), an explanation is provided, and the consequences can be accounted for in post-irradiation results. This work demonstrated that using subsized specimens for physical and thermal property measurements is suitable when necessary for fine-grain isotropic graphite.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Baseline Hypothetical Facility for the Production of 131 I and 99 Mo using Activation Targets

This report describes a hypothetical facility for production of medical radioisotopes via activation under the Proliferation Resistance and Optimization (PRO-X) program. The facility uses neutron activation of non-special nuclear material (SNM) to produce the medical isotopes 131 I and 99 Mo at a throughput of 60 Ci/week of 131 I and 5 Ci/week of 99 Mo. The hypothetical design was carried out using a 10 MWt research reactor. The precursors used for the activation process were TeO2 for 131 I and MoO 3 for 99 Mo. The processes are performed in 3 hot cells used for target receipt, extraction, purification low specific activity (LSA) generator introduction, and packaging. A fourth hotcell is used for waste processing. The hot cell processing area takes up a footprint of 15.4 m 2 with the total footprint of the facility, including space for administrative offices, non-rad labs, quality assurance, and radiation buffer areas set at 763 m 2 . Waste is produced at a weekly rate of 257.8 g low activity solid waste and 8032.7 mL of low activity liquid waste, 8032 mL of which is water. This baseline hypothetical facility for production of medical isotopes via activation was then compared and contrasted to the hypothetical facility for production of medical isotopes via fission products to show the differences in approach for the two production modes. The two production modes had several highlighted differences including the overall facility and hot cell layout, the type and amount of waste produced by the respective facilities, and economic factors impacting production mode. Finally, a decision tree for which production mode might be more beneficial for an entrant into medical isotope production was developed based on the differences examined and the desired output of medical isotopes desired by the entrant.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗