Search NASA⌕ Search

SEARCH · Search NASA

Results for “Isotope Program”

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 631 records · Page 35

Benchmark for Fuel Shuffling and Depletion for Pebble-Bed Reactors

Pebble bed reactors have specific operational characteristics when their fuel-cycle and fueling operations are considered. They are specifically distinguished by other type of nuclear reactor designs by their online fuel recycling scheme, where the fuel elements that have not yet reached discharge burnup can be reloaded and recycled continuously during normal operation. The fuel in a pebble bed reactor is not stationary and stochastically moves through the core once or several times during its lifetime, which allows them to operate without requiring a large excess reactivity hold for the burnup. However, this characteristic of pebble bed reactors introduces challenges in simulation, as each pebble can take many different trajectories through the core, its composition depends on the details of the irradiation history that is unique to its aggregated path through the core. For predicting the safety performance characteristics, such as source term, maximum fuel temperatures and fuel failure rates, etc., it is important to accurately incorporate the movement of pebbles through the core during their lifetime in a multi-physics simulation together with other phenomena. The equilibrium core analysis for pebble bed reactors are performed with multi-physics tools including fuel depletion in a multi pass reload coupled to the fuel movement. Currently, there are only a few legacy multi-physics simulation tools that can implement the pebble flow characteristics and perform equilibrium core analysis for pebble bed reactors. However, there are development efforts on-going under Department of Energy's Nuclear Energy Advanced Modelling and Simulation program and also in private industry for including these capabilities into their modelling and simulation tools. Any new development in the modelling and simulation tools needs to be validated by using tools such as experiments, analytical solutions or code-to-code benchmarks. In this work, a code-to-code benchmark for the equilibrium core analysis capability of pebble bed reactors was developed. Multiple cases were identified to capture different fuel cycle strategies that can be used in PBRs. The results of each case are presented in terms of overall equilibrium core characteristics: the discharge burnup; spatial burnup distribution; spatial isotopic distributions; axial and radial neutron flux distributions and power history of fuel elements per pass through the core for both a prototypical pebble bed High Temperature Gas-cooled Reactor and a prototypical pebble bed Fluoride-salt cooled High temperature Reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Computational Tool for Cosmic Ray Flux and Cosmogenic Activation Production Rates

Silicon charge couple devices (CCDs) are used in direct detection dark matter searches. Silicon can be obtained with very high purity and negligible amounts of radioactive contaminants. However, silicon exposed to cosmic rays can be activated, leading to the buildup of radioactive isotopes such as tritium. The decay of tritium can produce signals that cannot be distinguished from the expected signals of dark matter interactions and tritium produced by cosmic rays is expected as one of the leading sources of background for next-generation experiments. The cosmic ray flux depends on variables such as geomagnetic latitude and longitude, atmospheric depth/altitude, solar activity, and surrounding environment for example. The goal of this work is to create a tool that will evaluate the total cosmic ray exposure and related tritium production rate for any given exposure history of a silicon detector. EXcel-based Program for calculation Atmospheric Cosmic-ray Spectrum (EXPACS) was used to obtain the cosmic ray fluxes. We have worked on creating a 3-dimensional gird that spans latitude, longitude, and altitude, and can be used to interpolate the cosmic ray flux at any given location. This computation tool additionally has use for work requiring knowledge of the cosmic ray flux at multiple locations.

79 ASTRONOMY AND ASTROPHYSICS↗

Thermal Hydraulics Safety Analyses of TRIGA Reactor Upon Fuel Reconfiguration

Safety analyses of the Training, Research, Isotope, General Atomics (TRIGA) reactor at Kansas State University(KSU) is performed using thermal hydraulic simulations based on changes to reactivity or the volumetric heat generation rate due to changes in the positioning of the enriched fuel elements in the reactor core. If the reactivity or neutronics parameters are known, then they can be used to simulate the thermal energy state of the material which is crucial to assess the safety of the nuclear fuel inside a nuclear reactor. Several thermal hydraulics models and codes have been developed in the past to assess reactor safety. RELAP (Reactor Excursion and Leak Analysis Program) is one such tool widely used throughout the world for safety analyses. Although RELAP is considered a reliable tool by regulators, vendors, and operators, it lacks the numerical fidelity provided by 3-D computational fluid dynamics (CFD) models. This work involves the critical task of assessing safety of a research reactor upon fuel reconfiguration based on models developed in RELAP and comparing them with a commercially available 3-D CFD program. The TRIGA Mark-II reactor, designed by General Atomics and located at Kansas State University, is a class of small pool type inherently safe research reactor licensed to operate at steady state power of up to 1.25 MW. It has a cylindrical core measuring 0.45 meters diametrically and 0.38 meters axially positioned in a tank measuring 1.98 meters diametrically and 6.25 meters axially. It has uranium zirconium hydride (UZrH) fuel rods housed in stainless steel cladding material arranged in five concentric rings (B,C,D,E and F rings, in that order, moving outwards from the center). TRIGA fuel is nominally 20% enriched, but the reactivity of the fuel can be changed by using burnable poisons (not used at KSU), or by varying the total weight percent of uranium in the UZrH fuel meat. The TRIGA reactor is designed to be an inherently safe reactor and is designed to be cooled by natural convection. Thus it is of paramount importance to avoid the film boiling regime; failing to do so will subject the fuel to severe conditions, possibly resulting in its failure due to fuel/clad interactions. A high DNBR ratio is a primary consideration in the reactor core design and thus any modification in the core should ensure that this criteria is always met. In order to fully utilize the capabilities of KSU TRIGA reactor facility, it should be operated at its full capacity, i.e. 1.25 MW, which requires increasing the fuel loading in some of the E- and F-ring elements from 8.5% U (nominal) to 12.0% U (nominal). Only four elements are to be switched to 12% loading, causing local power peaking in these elements. The degree of power peaking to be expected in these 12%-loaded fuel elements was calculated based on a detailed MCNP (Monte Carlo N-Particle code) model of the reactor core. Some preliminary results of the calculations are presented in this paper based on changes in peaking factors and power levels. To predict reactor thermal behavior, axial changes in the peaking factors obtained from MCNP models were used to calculate volumetric energy density functions for a single channel one dimensional RELAP model and a three dimensional COMSOL model of the TRIGA MARK-II reactor. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Progress in the U.S./Japan PHENIX Project for the Technological Assessment of Plasma Facing Components for DEMO Reactors

The PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the response to high heat loads and tritium interactions of irradiated and unirradiated materials and components. This paper provides an overview of the progress achieved during the first 3 years and discusses the plan for the remainder of the project.

Kato, Yutai↗

Hydrogen Isotope MicroChemLab FY15

We have developed a new method to measure the composition of gaseous mixtures of any two hydrogen isotopes, as well as an inert gas component. When tritium is one of those hydrogen isotopes, there is usually some helium present, because the tritium decays to form helium at a rate of about 1% every 2 months. The usual way of measuring composition of these mixtures involves mass spectrometry, which involves bulky, energy-intensive, expensive instruments, including vacuum pumps that can quite undesirably disperse tritium. Our approach uses calorimetry of a small quantity of hydrogen-absorbing material to determine gas composition without consuming or dispersing the analytes. Our work was a proof of principle using a rather large and slow benchtop calorimeter. Incorporation of microfabricated calorimeters, such as those that have been developed in Sandia’s MicroChemLab program or that are now commercially available, would allow for faster measurements and a smaller instrument footprint.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Building Assessment of Radon Reduction Interventions with Energy Retrofits Expansion (The BEX Study): Final Report

This report presents the findings from a study that was performed to assess whether current precautionary measures used by the US Department of Energy Weatherization Assistance Program (WAP) are effective for preventing indoor radon increases following weatherization. The study results show that current practices have produced substantial benefit compared to previous practices, and that there are no statistically significant changes in indoor radon levels on the lowest living levels with these practices following weatherization.

07 ISOTOPE AND RADIATION SOURCES↗

Adaptive-in-temperature method for fast on-the-fly sampling of thermal neutron scattering data in MCNP6 [Slides]

The project objective and motivation is to develop thermal data libraries for selected isotopes in MCNP6 to support on-the-fly S(alpha, beta) sampling for temperature ranges applicable to nuclear criticality safety, thereby enhancing the physics treatment in MCNP6 so that it can perform fast on-the-fly sampling of S(alpha, beta) data at arbitrary temperature.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

UCO TRISO Minifuel FY23 NSUF-Kairos Power Post-Irradiation Examination Status Report

Irradiation of miniature tristructural isotropic (TRISO)–coated particle fuel compacts at high-power particle was performed in the Oak Ridge National Laboratory’s (ORNL’s) High Flux Isotope Reactor (HFIR) using the MiniFuel irradiation capability. Each compact comprised 20 TRISO particles with a low-enriched uranium carbide uranium oxide (UCO), natural UCO, or low-enriched UO 2 kernel within a graphitic matrix. After irradiation, the MiniFuel targets and subcapsules were disassembled to recover the irradiated fuel specimens and pursue post-irradiation examination (PIE) to inform Kairos Power on the fuel specimen performance. This report describes the PIE results collected to date, including dilatometry on the passive thermometry to confirm the irradiation temperature, fission gas release measurements, and gamma counting. This work was funded by the Nuclear Science User Facilities program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Software Verification of VARPOW

The VARPOW program is a post-processing utility program for DIF3D, specifically DIF3DVARIANT, and it was developed to provide interface files for the thermal analysis program DASSH. The basic methodology of VARPOW is to take the neutron and gamma flux (moments) calculated by DIF3D (or GAMSOR) and combine them with the heating (coefficient) cross sections to calculate the spatial power distributions using the DIF3DVARIANT spatial basis. VARPOW can use the output from GAMSOR (both steady state neutron and gamma flux calculations) or standard DIF3D/REBUS calculations (neutron flux only). The correct approach for defining the power distribution is to use GAMSOR as its purpose was to properly compute the gamma heating throughout the modeled domain. The power densities calculated by VARPOW are broken into fuel, cladding and coolant terms for which isotope-wise categorization is needed. VARPOW has built in options the user can select for the isotope categorization or VARPOW can import a file that details the isotope categorization. VARPOW can export the solution in the polynomial basis of DIF3D-VARIANT or the monomial basis of DIF3D-VARIANT. The purpose of this work is to verify the power distribution results calculated by VARPOW from both the GAMSOR and DIF3D input options and verify that the input and output options are consistent with the manual. Hand calculation and independent numerical calculation are used for this verification work.

97 MATHEMATICS AND COMPUTING↗

Software Verification of VARPOW

The VARPOW program is a post-processing utility program for DIF3D, specifically DIF3DVARIANT, and it was developed to provide interface files for the thermal analysis program DASSH. The basic methodology of VARPOW is to take the neutron and gamma flux (moments) calculated by DIF3D (or GAMSOR) and combine them with the heating (coefficient) cross sections to calculate the spatial power distributions using the DIF3DVARIANT spatial basis. VARPOW can use the output from GAMSOR (both steady state neutron and gamma flux calculations) or standard DIF3D/REBUS calculations (neutron flux only). The correct approach for defining the power distribution is to use GAMSOR as its purpose was to properly compute the gamma heating throughout the modeled domain. The power desnities calculated by VARPOW are broken into fuel, cladding and coolant terms for which isotope-wise categorization is needed. VARPOW has built in options the user can select for the isotope categorization or VARPOW can import a file that details the isotope categorization. VARPOW can export the solution in the polynomial basis of DIF3D-VARIANT or the monomial basis of DIF3D-VARIANT. The purpose of this work is to verify the power distribution results calculated by VARPOW from both the GAMSOR and DIF3D input options and verify that the input and output options are consistent with the manual. Hand calculation and independent numerical calculation are used for this verification work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Software Verification of VARPOW

The VARPOW program is a post-processing utility program for DIF3D, specifically DIF3DVARIANT, and it was developed to provide interface files for the thermal analysis program DASSH. The basic methodology of VARPOW is to take the neutron and gamma flux (moments) calculated by DIF3D (or GAMSOR) and combine them with the heating (coefficient) cross sections to calculate the spatial power distributions using the DIF3DVARIANT spatial basis. VARPOW can use the output from GAMSOR (both steady state neutron and gamma flux calculations) or standard DIF3D/REBUS calculations (neutron flux only). The correct approach for defining the power distribution is to use GAMSOR as its purpose was to properly compute the gamma heating throughout the modeled domain. The power desnities calculated by VARPOW are broken into fuel, cladding and coolant terms for which isotope-wise categorization is needed. VARPOW has built in options the user can select for the isotope categorization or VARPOW can import a file that details the isotope categorization. VARPOW can export the solution in the polynomial basis of DIF3D-VARIANT or the monomial basis of DIF3D-VARIANT. Recently the clad neutron damage calculation capability was introduced into VARPOW, and it now allows a pathway to get distributions of clad DPA. The purpose of this work is to verify the power distribution, fast neutron flux and clad DPA results calculated by VARPOW from both the GAMSOR and DIF3D input options and verify that the input and output options are consistent with the manual. Hand calculation and independent numerical calculation are used for this verification work.

Zhong, Zhaopeng [Argonne National Laboratory (ANL)↗

A New Process for Small-Batch Purification of the Medical Isotope Molybdenum-99: Non-Technical Overview

The U.S. medical community depends on a reliable supply of the radioisotope molybdenum-99 (Mo-99) for nuclear medical diagnostic procedures. Mo-99's decay product, technetium-99m (Tc-99m), is used in over 40,000 medical procedures in the United States each day to diagnose heart disease and cancer, to study organ structure and function, and to perform other important medical applications. For example, patients undergoing a common procedure—the cardiac “stress test”—likely have benefited from Tc-99m. Historically, Mo-99 was primarily produced through the fission of uranium-235, in the form of highly enriched uranium (HEU) targets irradiated in research and test reactors. HEU is a proliferation-sensitive material that, if diverted or stolen, could be used as a component of a nuclear weapon. NNSA’s Office of Material Management and Minimization (M3) manages the Molybdenum-99 (Mo-99) Program as part of its mission to minimize the use of HEU in civilian applications. The Mo-99 Program assists global Mo-99 production facilities in converting to non-HEU processes and supports the establishment of domestic supplies of Mo-99 without the use of proliferation-sensitive HEU. As part of this program, M3 funds U.S. national laboratories to provide non-proprietary technical support to U.S. companies working to establish non-HEU-based Mo-99 production capabilities. The results of this research are published on OSTI.gov for the benefit of the Mo-99 community and the public. However, it can be difficult for readers without a scientific background to understand and interpret these publications. In order to increase public understanding of the work being done in M3’s Mo-99 Program, this paper aims to provide an overview of a key, recent national laboratory technical publication in terms that can be understood by readers without a technical background. To accomplish this, the paper first explains key scientific concepts—primarily related to chemistry—that provide a foundation for understanding research in this area. This includes chromatography, absorption vs. adsorption, dissolution and precipitation, and liquid-liquid extraction. Drawing on these concepts, the paper then provides an explanation for non-technical audiences of the Argonne National Laboratory publication entitled Recovery of High Specific Activity Molybdenum-99 from Accelerator-Induced Fission on Low-Enriched Uranium for Technetium-99m Generators (Brown, M.A. et al., 2021) and related article Separation and Purification of Mo-99 Produced from Natural U3O8 Targets via Photo-Fission (Brown, M.A. et al., 2021).

Stamler, Bradley↗

pyEGAF: An open-source Python library for the Evaluated Gamma-ray Activation File

The Evaluated Gamma-ray Activation File (EGAF) is one of the most comprehensive resources for thermal neutron-capture data. This database contains data from prompt gamma activation analysis measurements carried out in a consistent manner using the same experimental configuration at the Budapest Research Reactor for 245 isotopes. Although these valuable datasets have been freely available for many years, one of the drawbacks is the outdated and cryptic Evaluated Nuclear Structure Data File (ENSDF) format that is currently adopted for dissemination, making it difficult for users unfamiliar with the format to access and utilize the data contained therein. Furthermore, the ENSDF format does not readily lend itself to modern computational technologies and a parser is required to interpret the complicated mixed-record format. To help overcome these challenges, we have developed a translator to convert the ENSDF-formatted datasets into an open standard JavaScript Object Notation (JSON) format enabling accessibility to applications using different programming languages running in different environments. To compliment this effort, we have also developed an open-source software package implemented in Python, pyEGAF, that is designed to interact with the JSON data structures for general purpose access, manipulation, and analysis of the neutron-capture $\gamma$-ray data in EGAF. The new format, together with the pyEGAF library, greatly enhances access to the wider applications community where EGAF data may be useful or is required.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Neutronic Safety Analysis of Pu-238 Production at Idaho National Laboratory

This analysis was completed to support the irradiation of plutonium fuel services (PFS) targets in the NEFT in the Advanced Test Reactor (ATR) as a part of the campaign to restart domestic production of plutonium-238 used in radioisotope power systems (RPS) by the National Aeronautical and Space Administration (NASA) and Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Nuclear Infrastructure Program (NE-3). Referred to as the PFS-ATR-GEN1-NEFT experiment, the assembly was designed to hold 46 PFS targets in the NEFT. The scope of this paper is to outline the MOPY method used to calculate the heat generation rates (HGRs), flux, fission density, and quantify the viability of the target design for Pu-238 production in ATR.

07 ISOTOPE AND RADIATION SOURCES↗

Maintaining a critical spectra within Monteburns for a gas-cooled reactor array by way of control rod manipulation

Our burnup study examined the effect of a predicted critical control rod position on the nuclide predictability of several axial and radial locations within a 4×4 graphite moderated gas cooled reactor fuel cluster geometry. To achieve this, a control rod position estimator (CRPE) tool was developed within the framework of the linkage code Monteburns between the transport code MCNP and depletion code CINDER90, and four methodologies were proposed within the tool for maintaining criticality. Two of the proposed methods used an inverse multiplication approach - where the amount of fissile material in a set configuration is slowly altered until criticality is attained - in estimating the critical control rod position. Another method carried out several MCNP criticality calculations at different control rod positions, then used a linear fit to estimate the critical rod position. The final method used a second-order polynomial fit of several MCNP criticality calculations at different control rod positions to guess the critical rod position. The results showed that consistency in prediction of power densities as well as uranium and plutonium isotopics was mutual among methods within the CRPE tool that predicted critical position consistently well. Finally, while the CRPE tool is currently limited to manipulating a single control rod, future work could be geared toward implementing additional criticality search methodologies along with additional features.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

An Analytic Benchmark for the Solution to the Isotopic Fission Spectrum Mixture Problem

In neutron transport calculations, it is common to specify material compositions in terms of constituent isotopes. Material compositions may be described by isotope number densities and associated microscopic cross sections. For general reaction cross sections, the macroscopic cross sections of a composition are simply the summation of the sum of the products of isotope number densities and microscopic cross sections. A notable exception is the mixture of the neutron fission spectrum in fissile material. To demonstrate proper and improper mixture of the neutron fission spectrum, a zero-dimensional test problem is developed. Using the test problem, it is demonstrated that the improper mixing of the fission spectrum results in an eigenvalue error of approximately 65 pcm. Though this may seem small, any error in such a simple calculation is unacceptable. In similar infinite-homogeneous test problems, eigenvalue errors of more than 300 pcm have been observed. It is also shown that the error due to fission spectrum mixing can be exaggerated for computer program verification purposes. It is concluded that the proper mixture of the neutron fission spectrum is essential for accurate neutron transport simulations. The effect of improper neutron fission spectrum mixture is demonstrated and quantified and this test problem may be used for verification purposes in the future.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Radioactive Material Inventory Optimization Using the Plutonium-239 Equivalent Curie Methodology

ABSTRACT The purpose of this paper is to provide the technical basis for application of the Plutonium 239 equivalent curies (PEC) concept used to optimize and limit the material-at-risk (MAR) in the Irradiated Materials Characterization Laboratory (IMCL) located at the Materials and Fuels Complex (MFC). The PEC concept offers a method to normalize all radionuclides to a common radiotoxic hazard index that can be summed to a single value. This paper describes the technical approach to applying the PEC concept, documents the weighting factors and specific activities to be utilized in the calculation, and provides an example of the application of the concept to a radionuclide distribution that is representative of materials encountered in the IMCL. The IMCL is operated to provide a dynamic, re-configurable capability to analyze and characterize irradiated or non-irradiated radioactive material samples. The operations unique to the IMCL include sample storage, preparation, and analysis using experiment-specific instrumentation such as, Electron probe micro-analyzer (EPMA), Focused ion beam (FIB), Transmission electron microscope (TEM), and Scanning electron microscope (SEM). The methodology used to characterize the radioactive material inventory enables the IMCL to operate in such a way that complements their capability and program goals without compromising the safety of the worker, co-located worker, or the general public. The PEC concept offers a method in which all radionuclide inventory is normalized to a common radiotoxic hazard index that can be summed to a single value. The PEC concept is useful in the comparison of various material types for hazard categorization and for identifying worst-case and bounding inventory.

07 ISOTOPE AND RADIATION SOURCES↗

Cr(VI) Occurrence and Geochemistry in Water From Public-Supply Wells in California

Hexavalent chromium, Cr(VI), in 918 wells sampled throughout California between 2004 and 2012 by the Groundwater Ambient Monitoring and Assessment Program—Priority Basin Project (GAMA—PBP) ranged from less than the study reporting limit (SRL) of 1 microgram per liter (μg/L) to 32 μg/L. Statewide, Cr(VI) was reported in 31 percent of sampled wells and equaled or exceeded the recently established (2014) California Maximum Contaminant Level (MCL) for Cr(VI) of 10 μg/L in 4 percent of sampled wells. Cr(VI) data collected for regulatory purposes overestimate Cr(VI) occurrence. Ninety percent of chromium was present as Cr(VI), which was detected more frequently and at higher concentrations in alkaline (pH > 8), oxic water, and more frequently in agricultural and urban land uses compared to native land uses. Chemical, isotopic (tritium and carbon-14), and noble-gas data show high Cr(VI) in water from wells in alluvial aquifers in the southern California deserts result from long groundwater-residence times and geochemical reactions such as silicate weathering that increase pH, while oxic conditions persist. High Cr(VI) in water from wells in alluvial aquifers along the west-side of the Central Valley results from high-chromium abundance in source rock eroded to form those aquifers, and areal recharge processes (including irrigation return) that mobilize chromium from the unsaturated zone. Cr(VI) co-occurred with oxyanions having similar chemistry, including vanadium, selenium, and uranium. Cr(VI) was positively correlated with nitrate, consistent with increased concentrations in areas of agricultural land use and mobilization of chromium from the unsaturated zone by irrigation return.

54 ENVIRONMENTAL SCIENCES↗