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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Fuel Cycle Research & Development: Technical Monthly – June 2012

Topics are: (1) MPACT Campaign - (a) Management and Integration • Coordination meetings between NE and NA-22, NA-24, and NA-82 were conducted the week of June 11th. Preparations are being made for the next MPACT working group meeting, scheduled for Aug 28-30 at Idaho Falls. In addition to covering accomplishments and discussing future plans, a site tour of INL facilities (MFC, EBR, ATR, INTEC) is being organized. (2) Accounting and Control Technologies - (a) Microcalorimetry • Now operating 256-pixel array at LANL. We are in the process of tuning detector parameters to improve and optimize performance. Preliminary measurements show approximate number of live pixels is similar to that observed previously at NIST. Continuing to study contribution to systematic error from uncertainties in tabulated gamma-ray energies. (b) Electrochemical Sensor • Testing of sensors fashioned from different precursor materials continued. SEM analysis of all used sensors has been or will be performed. (c) Lead Slowing Down Spectrometer • Ongoing perturbation calculations are providing information on the fundamental systematic error limits of LSDS. In order to achieve separating the contribution of Pu and 235U to the signal, there will need to be tight controls on systematic errors. Continuing to look into a He4 detector. Research into local construction of a He4 detector continued. We have started to apply the algorithm to test the LSDS using experimental data from previous RPI measurements. PNNL also developed a plan to address the lack of statistics in the MCNP modeling of the NGSI 64 assemblies. The ISU graduate student built and tested a fission chamber to gain experience with them. (d) Fast Neutron Imaging to Quantify Nuclear Materials • The imaging detector design was modified for each pixel to have an 8 x 8 pixel array. Quotations and purchasing process for components, including the new PSD scintillator are in progress. (e) Fast Neutron Multiplicity Analysis • The team submitted two papers to the upcoming INMM meeting that are related to the fast neutron multiplicity R&D effort. Progress was made on the project's main goal of designing a concept for a prototype fast-neutron multiplicity counter. We started laying out the outline for the final report. We have been working with our ORNL collaborators to develop a new digitizer system to support our experimental campaign planned for next year. (3) MPACT Analysis Tools - (a) Multi-isotope Process Monitor • Fuel characterization framework development continued during June. A report describing the methodologies is being completed. Kenneth Dayman, from University of Texas, spent a week at PNNL wrapping up his master's research and working on a journal submission covering that work. The target journal is the IEEE transactions on Nuclear Science; submission is planned for the end of July. A proposal to instrument H-Canyon is being prepared in conjunction with SRNL and the NNSA's NGSI program. The impact of gamma-ray spectrum counting statistics on the precision of relative radioisotope component intensities as reconstructed via Principal Component Regression (PCR) continued in June with Monte Carlo simulations of a two-component (i.e., two radioisotope) system. This work generalizes earlier studies in FY12 in which Poisson counting variations of only a single spectrum component were simulated. (b) Modeling and Simulation for Analysis of Safeguards Performance (Electrochemical) • Preliminary insights into safeguards challenges and the initial design for an electrochemical plant have been written up into an INMM paper and will be presented at the INMM Summer Meeting. Work is currently adding a new visualization capability for integrating materials accountancy with physical protection. (c) Material Control including Process Monitoring (Pattern Recognition, Sensors) • Fabrication of quartz chips continued at an external foundry. Awaiting delivery of the heat exchange manifold and chip holder. (d) MPACT System Integration and Technical Support • The initial report on cost-basis metrics for nuclear material security risk and proliferation risk was completed as scheduled. (4) Safeguards and Security by Design - (a) Used fuels storage security analysis, guidance and best practices • RFP was finalized and sent out. Awaiting response from World Institute for Nuclear Security. Coordinated with Scott Demuth on joint report • LANL milestone on best practices. Attended Nuclear Energy Institute National Nuclear Security Conference. Developed plan for soliciting input on prioritized issues for used fuel storage security. Worked on setting up a meeting with the Nuclear Regulatory Commission. Worked on FY2013 work package planning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of an Efficient Approach to Perform Neutronics Simulations for Plutonium-238 Production

Conversion of 238Pu decay heat into usable electricity is imperative to power National Aeronautics and Space Administration (NASA) deep space exploration missions; however, the current stockpile of 238Pu is diminishing and the quality is less than ideal. In response, the US Department of Energy and NASA have undertaken a program to reestablish a domestic 238Pu production program and a technology demonstration sub-project has been initiated. Neutronics simulations for 238Pu production play a vital role in this project because the results guide reactor safety-basis, target design and optimization, and post-irradiation examination activities. A new, efficient neutronics simulation tool written in Python was developed to evaluate, with the highest fidelity possible with approved tools, the time-dependent nuclide evolution and heat deposition rates in 238Pu production targets irradiated in the High Flux Isotope Reactor (HFIR). The Python Activation and Heat Deposition Script (PAHDS) was developed specifically for experiment analysis in HFIR and couples the MCNP5 and SCALE 6.1.3 software quality assured tools to take advantage of an existing high-fidelity MCNP HFIR model, the most up-to-date ORIGEN code, and the most up-to-date nuclear data. Three cycle simulations were performed with PAHDS implementing ENDF/B-VII.0, ENDF/B-VII.1, and the Hybrid Library GPD-Rev0 cross-section libraries. The 238Pu production results were benchmarked against VESTA-obtained results and the impact of various cross-section libraries on the calculated metrics were assessed.

depletion↗
DOE OSTI2025

IER 516 - Zirconium Test Assembly (ZTA) Preliminary Experimental Design [Slides]

Zirconium present in large quantities in criticality safety applications such as fuel manufacturing and processing facilities, and in commercial thermal nuclear reactors. ICSBEP lacks benchmarks optimized for zirconium sensitivity to validate zirconium evaluations. Evaluators lack isotopic differential nuclear data in regions of high importance to fission systems.

K-PSO 2.0↗
DOE OSTI2013

Dismantling the nuclear research reactor Thetis

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written). (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Constraining Genome-Scale Models to Represent the Bow Tie Structure of Metabolism for 13 C Metabolic Flux Analysis

Determination of internal metabolic fluxes is crucial for fundamental and applied biology because they map how carbon and electrons flow through metabolism to enable cell function. 13 C Metabolic Flux Analysis ( 13 C MFA) and Two-Scale 13 C Metabolic Flux Analysis (2S- 13 C MFA) are two techniques used to determine such fluxes. Both operate on the simplifying approximation that metabolic flux from peripheral metabolism into central “core” carbon metabolism is minimal, and can be omitted when modeling isotopic labeling in core metabolism. The validity of this “two-scale” or “bow tie” approximation is supported both by the ability to accurately model experimental isotopic labeling data, and by experimentally verified metabolic engineering predictions using these methods. However, the boundaries of core metabolism that satisfy this approximation can vary across species, and across cell culture conditions. Here, we present a set of algorithms that (1) systematically calculate flux bounds for any specified “core” of a genome-scale model so as to satisfy the bow tie approximation and (2) automatically identify an updated set of core reactions that can satisfy this approximation more efficiently. First, we leverage linear programming to simultaneously identify the lowest fluxes from peripheral metabolism into core metabolism compatible with the observed growth rate and extracellular metabolite exchange fluxes. Second, we use Simulated Annealing to identify an updated set of core reactions that allow for a minimum of fluxes into core metabolism to satisfy these experimental constraints. Together, these methods accelerate and automate the identification of a biologically reasonable set of core reactions for use with 13 C MFA or 2S- 13 C MFA, as well as provide for a substantially lower set of flux bounds for fluxes into the core as compared with previous methods. We provide an open source Python implementation of these algorithms at https://github.com/JBEI/limitfluxtocore.

13C metabolic flux analysis↗
DOE OSTI2018

Reconstruction of Irradiation History of Am-241/Be Startup Sources for Disposal at WIPP - 18644

The National Nuclear Security Administration, Global Material Security (GMS)/ Off-site Source Recovery Program (OSRP) requires an approved final disposition pathway for sealed neutron startup sources with undocumented or questionable irradiation histories. To obtain approval for WIPP disposal, a reliable method is required to determine yields of transmutation, as well as fission and activation products contained within the source at the time of disposal. This paper presents a method to reconstruct the irradiation history of americium beryllium (Am/Be) reactor startup sources using gamma spectroscopy, ORIGEN modeling and the initial composition of the source. Using this method, measured isotopes in the Am/Be startup sources are related to the WIPP disposal requirements. Maximum allowable measurement uncertainties have been quantified and related to uncertainties in the WIPP required isotopes. Characterization calculations and experimental measurements of irradiated Am/Be sources are discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Communication: Localized molecular orbital analysis of the effect of electron correlation on the anomalous isotope effect in the NMR spin-spin coupling constant in methane

We discuss the effect of electron correlation on the unexpected differential sensitivity (UDS) in the {sup 1}J(C–H) coupling constant of CH{sub 4} using a decomposition into contributions from localized molecular orbitals and compare with the {sup 1}J(N–H) coupling constant in NH{sub 3}. In particular, we discuss the well known fact that uncorrelated coupled Hartree-Fock (CHF) calculations are not able to reproduce the UDS in methane. For this purpose we have implemented for the first time a localized molecular orbital analysis for the second order polarization propagator approximation with coupled cluster singles and doubles amplitudes—SOPPA(CCSD) in the DALTON program. Comparing the changes in the localized orbital contributions at the correlated SOPPA and SOPPA(CCSD) levels and at the uncorrelated CHF level, we find that the latter overestimates the effect of stretching the bond between the coupled atoms on the contribution to the coupling from the localized bonding orbital between these atoms. This disturbs the subtle balance between the molecular orbital contributions, which lead to the UDS in methane.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2012

Kinetic and structural studies, origins of selectivity, and interfacial charge transfer in the artificial photosynthesis of CO

The effective design of an artificial photosynthetic system entails the optimization of several important interactions. Herein we report stopped-flow UV-Vis spectroscopy, X-ray crystallography, DFT, and electrochemical kinetic studies of the Re(bipy-tBu)(CO)3(L) catalyst system. A remarkable selectivity for CO2 over H+ was observed by stopped-flow UV-Vis spectroscopy of [Re(bipy-tBu)(CO)3]-. The pseudo-first order rate constant for the reaction with 10 mM CO2 in THF is 35 s-1. This is ca. 15-20 times faster than the reactions with water or methanol at the same concentration in THF. X-ray crystallography and DFT studies of the doubly-reduced anionic species suggest that the HOMO has mixed metal-ligand character rather than being purely dz 2, which is thought to aid catalytic selectivity by favoring binding of CO2 over H+. Electrocatalytic studies performed with the addition of Brönsted acids reveal a primary H/D kinetic isotope effect, indicating that transfer of protons to Re-CO2 is involved in the rate limiting step. Lastly, the effects of electrode surface modification on interfacial electron transfer between a semiconductor and catalyst were investigated and found to affect the observed catalytic rates up to seven-fold, indicating that the properties of the electrode surface should not be overlooked when developing a homogeneous artificial photosynthetic system. This research was supported at the University of Washington, Seattle by the Camille and Henry Dreyfus Postdoctoral Program in Environmental Chemistry (for a fellowship to A. J. M. M.), and, for funds to purchase the stopped-flow instrument, the U.S. National Institutes of Health 13 (Grant GM-50422 to JMM), and the Center for Molecular Electrocatalysis, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences.

Smieja, Jonathan M↗
DOE OSTI2024

Krypton-85 chronometry of spent nuclear fuel

We describe the use of the radionuclide 85 Kr, which is produced by nuclear fission and has a half-life of 10.76 years, to determine the age of spent nuclear fuel. The method is based on mass-spectrometric measurement of the relative abundance of fissiogenic Kr isotopes extracted from a fuel sample, and we show that it can be applied to micron-scale particles of spent fuel that are analogous to particles that have been released into the environment from various nuclear facilities in the past. 85 Kr chronometry is potentially valuable for identifying and attributing nuclear materials, grouping samples into collections of common origin that can be used to reconstruct the origin and irradiation history of the material, and verifying declared nuclear activities in the context of international monitoring programs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2 ) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

The multi-scattering model for calculations of positron spatial distribution in the multilayer stacks, useful for conventional positron measurements

The spatial distribution of positrons emitted from radioactive isotopes into stacks or layered samples is a subject of the presented report. It was found that Monte Carlo (MC) simulations using GEANT4 code are not able to describe correctly the experimental data of the positron fractions in stacks. The mathematical model was proposed for calculations of the implantation profile or positron fractions in separated layers or foils being components of a stack. The model takes into account only two processes, i.e., the positron absorption and backscattering at interfaces. The mathematical formulas were applied in the computer program called LYS-1 (layers profile analysis). The theoretical predictions of the model were in the good agreement with the results of the MC simulations for the semi infinite sample. The experimental verifications of the model were performed on the symmetrical and non-symmetrical stacks of different foils. The good agreement between the experimental and calculated fractions of positrons in components of a stack was achieved. Also the experimental implantation profile obtained using the depth scanning of positron implantation technique is very well described by the theoretical profile obtained within the proposed model. The LYS-1 program allows us also to calculate the fraction of positrons which annihilate in the source, which can be useful in the positron spectroscopy.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Coupled Decay Heat and Thermal Hydraulic Capability for Loss-of-Coolant Accident Simulations

As the nuclear energy industry considers ways to achieve improved economics in the current fleet of light-water reactors (LWRs), one possible approach is to operate each cycle for longer durations. This causes a greater portion of the fuel to be burned and reduces the frequency of outages, which ultimately reduces the cost to operate the reactor. However, this also leads to higher burnup fuels than has traditionally been allowed in these reactors. Thus, there are concerns about integrity of high-burnup (HBu) fuel, especially during accident conditions such as loss-of-coolant accidents (LOCAs), as shown by Capps et al.. To investigate these concerns, advanced modeling and simulation capabilities are being leveraged to determine the susceptibility of HBu fuel to fuel fragmentation, relocation, and dispersion (FFRD). Improvements have previously been made to fuel performance capabilities to more accurately model these phenomena; multiphysics simulations have also been conducted to determine the power and burnup histories of the HBu fuel, which are needed as inputs for the fuel performance calculations. Most recently, new statistical approaches have been developed to identify a subset of fuel rods that have greater FFRD susceptibility, reducing the total number of fuel performance simulations required. Prior LOCA simulations have relied on the TRACE systems code, which can model the core and primary loop during accident conditions. TRACE includes many models for various aspects of the primary loop, but two sets of models are important for this report. First, TRACE uses a lumped-fuel approach for modeling the core. This approximates the ~50,000 fuel rods in the core with a much smaller number of rods. The rods can be lumped in various ways as determined by the user. For example, one lumped rod may be used to represent all rods in an assembly, sometimes with an additional rod representing the hottest fuel rod. However, due to runtime constraints and complexity of modeling, a more common approach is to group several assemblies or larger regions of the core into single lumped rods. These lumping schemes apply not only to fuel rods but to flow channels as well. Second, TRACE has several different models for treating decay heat, ranging from pregenerated decay heat curves based on an ANSI/ANS-5.1 standard (hereinafter abbreviated simply as ANSI) to explicit time-dependent heat inputs from the user. None of these models account for differences in isotopics between different rods, which is an approximation the work in this report seeks to eliminate. This report focuses on the implementation of coupled decay heat capabilities in the Virtual Environment for Reactor Applications (VERA) code suite to address a gap identified in previous LOCA simulations. This constitutes an improvement for both the lumped-fuel and decay heat models in TRACE. VERA has been developed to perform high-fidelity, whole-core multiphysics simulations for LWRs. Previously, during the Consortium for Advanced Simulation of LWRs (CASL) program, the emphasis was on providing accurate steady-state analysis—with a secondary focus on reactivity insertion accident (RIA) analysis—to address operational challenges in the nuclear energy industry. Under the Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, these capabilities are being extended to a broader range of transient analyses with the goal of quantifying the risk of fuel failures such as FFRD. To properly model such conditions with VERA, decay heat calculations have been integrated with the multiphysics to enable rod-by-rod thermal hydraulic (TH) conditions to be driven by the decay heat in long-running accidents such as LOCAs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

NEW MATERIALS DEVELOPED TO MEET REGULATORY AND TECHNICAL REQUIREMENTS ASSOCIATED WITH IN-SITU DECOMMISSIONING OF NUCLEAR REACTORS AND ASSOCIATED FACILITIES

For the 2010 ANS Embedded Topical Meeting on Decommissioning, Decontamination and Reutilization and Technology, Savannah River National Laboratory's Mike Serrato reported initial information on the newly developed specialty grout materials necessary to satisfy all requirements associated with in-situ decommissioning of P-Reactor and R-Reactor at the U.S. Department of Energy's Savannah River Site. Since that report, both projects have been successfully completed and extensive test data on both fresh properties and cured properties has been gathered and analyzed for a total of almost 191,150 m{sup 3} (250,000 yd{sup 3}) of new materials placed. The focus of this paper is to describe the (1) special grout mix for filling the P-Reactor vessel (RV) and (2) the new flowable structural fill materials used to fill the below grade portions of the facilities. With a wealth of data now in hand, this paper also captures the test results and reports on the performance of these new materials. Both reactors were constructed and entered service in the early 1950s, producing weapons grade materials for the nation's defense nuclear program. R-Reactor was shut down in 1964 and the P-Reactor in 1991. In-situ decommissioning (ISD) was selected for both facilities and performed as Comprehensive Environmental Response, Compensations and Liability Act actions (an early action for P-Reactor and a removal action for R-Reactor), beginning in October 2009. The U.S. Department of Energy concept for ISD is to physically stabilize and isolate intact, structurally robust facilities that are no longer needed for their original purpose of producing (reactor facilities), processing (isotope separation facilities), or storing radioactive materials. Funding for accelerated decommissioning was provided under the American Recovery and Reinvestment Act. Decommissioning of both facilities was completed in September 2011. ISD objectives for these CERCLA actions included: (1) Prevent industrial worker exposure to radioactive or hazardous contamination exceeding Principal Threat Source Material levels; (2) Minimize human and ecological exposure to unacceptable risk associated with radiological and hazardous constituents that are or may be present; (3) Prevent to the extent practicable the migration of radioactive or hazardous contaminants from the closed facility to the groundwater so that concentrations in groundwater do not exceed regulatory standards; (4) Eliminate or control all routes of human exposure to radiological and chemical contamination; and (5) Prevent animal intruder exposure to radioactive and hazardous contamination.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Approximating the r-Process on Earth with Thermonuclear Explosions. Lessons Learned and Unanswered Questions

During the astrophysical r-process, multiple neutron captures occur so rapidly on target nuclei that their daughter nuclei generally do not have time to undergo radioactive decay before another neutron is captured. The r-process can be approximately simulated on Earth in certain types of thermonuclear explosions through an analogous process of rapid neutron captures known as the "prompt capture" process. Between 1952 and 1969, 23 nuclear tests were fielded by the US which were involved (at least partially) with the "prompt capture" process. Of these tests, 15 were at least partially successful. Some of these tests were conducted under the Plowshare Peaceful Nuclear Explosion Program as scientific research experiments. It is now known that the USSR conducted similar nuclear tests during 1966 to 1979. The elements einsteinium and fermium were first discovered by this process. The most successful tests achieved 19 successive neutron captures on the initial target nuclei. A review of the US program, target nuclei used, heavy element yields, scientific achievements of the program, and how some of the results have been used by the astrophysical community is given. Finally, some unanswered questions concerning very neutron-rich nuclei that could potentially have been answered with additional nuclear experiments is presented.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

Characterization of gamma-ray background outside of the High Flux Isotope Reactor

Static gamma-ray measurements collected between September 2017 and March 2018 in the vicinity of the High Flux Isotope Reactor (HFIR) and the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory are presented. Temporal events are identified and tracked, including radioactive material transfers, rain events, and 41 Ar emissions. The remaining data is used to determine the average background and its variation. Finally, these results will be used to benchmark Monte Carlo transport models of the HFIR/REDC vicinity and provide training data for material transfer tracking algorithms and data fusion algorithms to determine HFIR operational status under the Multi-Informatics for Nuclear Operations Scenarios program.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Lipid Informed Quantitation and Identification

LIQUID (Lipid Informed Quantitation and Identification) is a software program that has been developed to enable users to conduct both informed and high-throughput global liquid chromatography-tandem mass spectrometry (LC-MS/MS)-based lipidomics analysis. This newly designed desktop application can quickly identify and quantify lipids from LC-MS/MS datasets while providing a friendly graphical user interface for users to fully explore the data. Informed data analysis simply involves the user specifying an electrospray ionization mode, lipid common name (i.e. PE(16:0/18:2)), and associated charge carrier. A stemplot of the isotopic profile and a line plot of the extracted ion chromatogram are also provided to show the MS-level evidence of the identified lipid. In addition to plots, other information such as intensity, mass measurement error, and elution time are also provided. Typically, a global analysis for 15,000 lipid targets.

Crowell, Kevin↗
DOE OSTI2018

PUMA: Plutonium Modeling and Analysis Theory Manual Rev 1.0

The high rate of alpha decay of the plutonium-238 isotope lends the material to be used for many unique space applications where the reach of solar radiation is either too weak or unavailable. With a power density of 0.54 watts per gram, plutonium-238 has been used to power space exploration satellites from mars rovers to deep space probes. Los Alamos National Laboratory (LANL) has remained the primary location for fabrication of the plutonium oxide pellets. In an effort to apply modern techniques and tools towards supporting LANL’s fabrication responsibilities, PUMA was created under an LDRD funded program to gain understanding of plutonium oxide heat source pellets during fabrication.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗