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At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

The Use of Subsurface Barriers to Support Treatment of Metals and Reduce the Flux of Tritium to Fourmile Branch at the Savannah River Site in South Carolina - 13358

The Savannah River Site (SRS) produced tritium, plutonium, and special nuclear materials for national defense, medicine, and the space programs. Acidic groundwater plumes containing metals, metallic radionuclides, non-metallic radionuclides and tritium sourced from the F and H Area Seepage Basins have impacted the surface water of Fourmile Branch on SRS. Tritium releases from Fourmile Branch have impacted the water quality within areas of the Savannah River adjacent to the SRS, and this circumstance has been an ongoing regulatory concern. The F and H Area Seepage Basins operated until 1988 for the disposition of deionized acidic waste water from the F and H Separations Facilities. The waste water contained dilute nitric acid and low concentrations of non-radioactive metals, and radionuclides, with the major isotopes being Cs-137, Sr-90, U-235, U-238, Pu-239, Tc-99, I-129, and tritium. The tritium concentration in the waste water was relatively elevated because there is not a practicable removal method in water. The acid content of the waste water during the operational period of the basins was equal to 12 billion liters of nitric acid. The seepage basins were closed in 1988 and backfilled and capped by 1991. The plumes associated with the F and H basins cover an area of nearly 2.4 square kilometers (600 acres) and discharge along ∼2,600 meters of Fourmile Branch. The acidic nature of the plumes and their overall discharge extent along the branch represent a large challenge with respect to reducing contaminant flux to Fourmile Branch. The introduction of nitric acid into the groundwater over a long time effectively reduced the retardation of metal migration from the basins to the groundwater and in the groundwater to Fourmile Branch, because most negatively charged surfaces on the aquifer materials were filled with hydrogen ion. Two large pump and treat systems were constructed in 1997 and operated until 2003 in an attempt to capture and control the releases to Fourmile Branch. The operating cost, including waste disposal, for the two systems was ∼$1.3 M/month. Both systems employed reinjection of tritiated water up gradient of the extraction, and produced large quantities of waste from non-tritium isotopes and metals removal prior to reinjection. Both systems were determined to be ineffective and potentially detrimental with respect to limiting the flux of contaminants to Fourmile Branch. After it became apparent that there was very little benefit to continued operation of the systems, and the staggering cost of operations was recognized by the SRS and regulators, a new remedy was developed. The new system uses vertical subsurface barriers to redirect groundwater flow to limit the transport of contaminants to the stream. The barriers were constructed of acid resistant grout using deep soil mixing techniques. The grout mixture used low swelling clay, fly ash, and sodium hydroxide to form a pozzolana material with low permeability and low strength. The SRS and regulators agreed to a series of remedial goals, with the first goal to reduce tritium flux to the stream by 70% and bring constituents other than tritium to groundwater protection standards. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

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↗
DOE OSTI2013

Tritium permeation experiments using reduced activation ferritic/martensitic steel tube and erbium oxide coating

Low concentration tritium permeation experiments have been performed on uncoated F82H and Er2O3-coated tubular samples in the framework of the Japan-US TITAN collaborative program. Tritium permeability of the uncoated sample with 1.2 ppm tritium showed one order of magnitude lower than that with 100% deuterium. The permeability of the sample with 40 ppm tritium was more than twice higher than that of 1.2 ppm, indicating a surface contribution at the lower tritium concentration. The Er2O3-coated sample showed two orders of magnitude lower permeability than the uncoated sample, and lower permeability than that of the coated plate sample with 100% deuterium. It was also indicated that the memory effect of ion chambers in the primary and secondary circuits was caused by absorption of tritiated water vapor that was generated by isotope exchange reactions between tritium and surface water on the coating.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

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↗
DOE OSTI2012

Ambient Monitoring for Sinclair and Dyes Inlets, Puget Sound, Washington: Chemical Analyses for 2012 Regional Mussel Watch

Under the Project ENVVEST Final Project Agreement, the Puget Sound Naval Shipyard & Intermediate Maintenance Facility (PSNS&IMF), Environmental Protection Agency (EPA), Washington State Department of Ecology (Ecology), and local stakeholders have worked collaboratively to improve the environmental quality of Sinclair and Dyes Inlets. A regional mussel monitoring program began in 2010 to assess the status and trend of ecological resources, assess the effectiveness of cleanup and pollution control measures, and determine if discharges from all sources are protective of beneficial uses including aquatic life. The program collected indigenous mussels to represent a time-integrated measure of bioavailable metals and organic chemicals present in the water column. This document supplements the 2010 indigenous mussel data with 2012 data to provide two years of data on the chemical residue of mussels present in the inter-tidal regions of Sinclair Inlet, Dyes Inlet, Port Orchard Passage, Rich Passage, Agate Passage, Liberty Bay, and Keyport Lagoon. The 2012 data set added one station at PSNS&IMF and one market samples from Penn Cove. Indigenous mussels were collected from a small boat and/or from along the shoreline, measured, composited, and analyzed for percent lipids, percent moisture, stable isotopes of carbon and nitrogen, and a suite of trace metals and organic contaminants. The trace metals included silver (Ag), arsenic (As), cadmium (Cd), copper (Cu), mercury (Hg), lead (Pb), and zinc (Zn). The organic contaminants included the list of NOAA Status and Trends 20 polychlorinated biphenyls (PCB) congeners and suite of parent and methylated polycyclic aromatic hydrocarbons (PAHs). The average lengths between the 2010 and 2012 data were generally less than 30% relative percent difference (RPD). Generally, the metals concentrations were lower in 2012 than 2010 with some notable exceptions in Sinclair Inlet and Rich Passage where increases in Ag, Hg, Pb, Cu, and Zn exceeded an RPD of 50% between years. However, they did not exceed the bioaccumulation critical values or the critical body residues corresponding to the no observed effect dose (NOED) and the lowest observed effect dose (LOED) with one exception. The Cd concentrations exceeded the NOED and LOED for the Manchester Lab Pier and the Pike Place Market samples. For the PAHs and PCB, the 2012 data were generally lower than 2010 and some cases significantly lower for PAHs and none of the available invertebrate benchmarks were exceeded.

Brandenberger, Jill M.↗
DOE OSTI2022

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↗
DOE OSTI2016

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↗
DOE OSTI2020

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↗
DOE OSTI2012

Definition of parameters for quality assurance of flattening filter free (FFF) photon beams in radiation therapy

Purpose: Flattening filter free (FFF) beams generated by medical linear accelerators have recently started to be used in radiotherapy clinical practice. Such beams present fundamental differences with respect to the standard filter flattened (FF) beams, making the generally used dosimetric parameters and definitions not always viable. The present study will propose possible definitions and suggestions for some dosimetric parameters for use in quality assurance of FFF beams generated by medical linacs in radiotherapy. Methods: The main characteristics of the photon beams have been analyzed using specific data generated by a Varian TrueBeam linac having both FFF and FF beams of 6 and 10 MV energy, respectively. Results: Definitions for dose profile parameters are suggested starting from the renormalization of the FFF with respect to the corresponding FF beam. From this point the flatness concept has been translated into one of 'unflatness' and other definitions have been proposed, maintaining a strict parallelism between FFF and FF parameter concepts. Conclusions: Ideas for quality controls used in establishing a quality assurance program when introducing FFF beams into the clinical environment are given here, keeping them similar to those used for standard FF beams. By following the suggestions in this report, the authors foresee that the introduction of FFF beams into a clinical radiotherapy environment will be as safe and well controlled as standard beam modalities using the existing guidelines.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

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↗
DOE OSTI2015

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↗
DOE OSTI2013

Characterization of Cu-SSZ-13 NH3 SCR Catalysts: an in situ FTIR Study

The adsorption of CO and NO over Cu-SSZ-13 zeolite catalysts, highly active in the selective catalytic reduction of NOx with NH3, was investigated by FTIR spectroscopy, and the results obtained were compared to those collected from other Cu-ion exchanged zeolites (Y,FAU and ZSM-5). At low CO pressures at room temperature (295 K) CO form monocarbonyls exclusively on the Cu+ ions, while in the presence of gas phase CO dicarbonyls on Cu+ and adsorbed CO on Cu2+ centers form, as well. At low (cryogenic) sample temperatures tricarbonyl formation on Cu+ sites was also observed. The adsorption of NO produces IR bands that can be assigned to nitrosyls bound to both Cu+ and Cu2+ centers, and NO+ species located in charge compensating cationic positions of the chabasite framework. On the reduced Cu-SSZ-13 samples the formation of N2O was also detected. The assignment of the adsorbed NOx species was aided by adsorption experiments with isotopically labeled 15NO. The movement of Cu ions from the sterically hindered six member ring position to the more accessible cavity positions as a result of their interaction with adsorbates (NO and H2O) was clearly evidenced. Comparisons of the spectroscopy data obtained in the static transmission IR system to those collected in the flow-through diffuse reflectance cell points out that care must be taken when conclusions are drawn about the adsorptive and reactive properties of metal cation centers based on a set of data collected under well defined, specific experimental conditions. Financial support was provided by the US Department of Energy (DOE), Office of Energy Efficiency and Renewable Energy, Vehicle Technologies Program. This work was performed in the Environmental Molecular Sciences Laboratory (EMSL) at the Pacific Northwest National Laboratory (PNNL). The EMSL is a national scientific user facility supported by the US DOE, Office of Biological and Environmental Research. PNNL is a multi-program national laboratory operated for the US DOE by Battelle.

adsorption of CO and NO↗
DOE OSTI2017

Two-Dimensional Neutronic and Fuel Cycle Analysis of the Transatomic Power Molten Salt Reactor

This status report presents the results from the first phase of the collaboration between Transatomic Power Corporation (TAP) and Oak Ridge National Laboratory (ORNL) to provide neutronic and fuel cycle analysis of the TAP core design through the Department of Energy Gateway for Accelerated Innovation in Nuclear, Nuclear Energy Voucher program. The TAP design is a molten salt reactor using movable moderator rods to shift the neutron spectrum in the core from mostly epithermal at beginning of life to thermal at end of life. Additional developments in the ChemTriton modeling and simulation tool provide the critical moderator-to-fuel ratio searches and time-dependent parameters necessary to simulate the continuously changing physics in this complex system. Results from simulations with these tools show agreement with TAP-calculated performance metrics for core lifetime, discharge burnup, and salt volume fraction, verifying the viability of reducing actinide waste production with this design. Additional analyses of time step sizes, mass feed rates and enrichments, and isotopic removals provide additional information to make informed design decisions. This work further demonstrates capabilities of ORNL modeling and simulation tools for analysis of molten salt reactor designs and strongly positions this effort for the upcoming three-dimensional core analysis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Candidacy Renewal Report for ICERR Designation

The International Atomic Energy Agency (IAEA) has established a designation for an International Centre based on Research Reactors (ICERR). The intention of this designation is to provide a vehicle for IAEA member states to access international research-reactor and ancillary nuclear research and development infrastructure. The U.S. Department of Energy (DOE) has made a commitment to world leadership in the development of advanced nuclear energy, science, and technology. To this end, DOE has established programs and initiatives to enhance this leadership role. ICERR designation constitutes an important step in achieving the DOE vision. DOE, represented by Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL), is submitting this application for re-designation as an IAEA ICERR. Both INL and ORNL have a decades-long and storied history that supports nuclear research, development, and deployment both nationally and internationally. Both have a history of safe and efficient nuclear operations and have demonstrated a track record of international collaboration and cooperation. INL and ORNL are home to two primary and two secondary research reactors. The primary reactors are Advanced Test Reactor at INL, and High-Flux Isotope Reactor at ORNL. Secondary reactors include: Neutron Radiography Reactor at INL, Transient Reactor Test (TREAT) Facility at INL, and INL Advanced Test Reactor: Critical Facility (ATRC). ICERR designation will also include numerous ancillary facilities at both laboratories, including the following: Multiple post-irradiation examination facilities at both INL and ORNL, Radiochemistry hot-cell facilities at ORNL, INL nuclear-fuel-fabrication and materials-science facilities, and INL and ORNL radiological and nuclear analytical capabilities.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Improved $^{95}\mathrm{Mo}$ neutron resonance parameters and astrophysical reaction rates

We report improved 95 Mo neutron resonance parameters and reaction rates are important for nuclear astrophysics, testing nuclear models, and nuclear criticality safety. However, despite many previous neutron-capture and total cross-section measurements on this nuclide, there still is much room for improvement as well as several discrepancies. For example, there are very few firm resonance spin and parity assignments; average resonance parameters are available only for each parity, the currently recommended astrophysical reaction rate results in disagreements between stellar models and meteoric isotopic anomalies, and there are substantial disagreements in the neutron-capture cross section at low energies important for nuclear criticality safety. To obtain an improved set of neutron resonance parameters and astrophysical reaction rates for 95 Mo. High-resolution neutron-capture and transmission data were measured at the Oak Ridge Electron Linear Accelerator (ORELA) using highly isotopically enriched 95 Mo samples. The neutron-capture apparatus, data reduction, and analysis were improved so that information contained in the γ-ray cascade following neutron capture were used to assign resonance J π values. Following this, simultaneous analysis of the new neutron-capture and transmission data was used to obtain resonance energies, gamma widths, and neutron widths and their uncertainties to a maximum energy of 10 keV. Accurate neutron-capture cross sections also were obtained for the unresolved resonance region to a maximum energy of 500 keV and, together with the new resonance parameters, used to calculate the astrophysical reaction rates in the temperature range from 5 to 30 keV. A vastly improved set of 95Mo neutron resonance parameters and an astrophysical reaction rate accurate to about 3% were obtained. Firm J π assignments were determined for 261 of the 314 observed resonances. This is a very large improvement over the previously published 32 firm J π assignments for 108 resonances. Also, the number of resonances having both firm J π assignments and Γ γ values was increased by almost a factor of 24—from 11 to 261. Neutron- and total-radiation-width distributions and average resonance spacings, average total radiation widths, and neutron strength functions were obtained for the six different s- and p-wave possibilities. Parameters for the lowest s-wave resonance, which is most important for criticality benchmarks, were obtained with high accuracy. Simple modification of the neutron-capture apparatus and expansion and improvement of data analysis techniques led to a large increase in firm J π assignments for 95 Mo neutron resonances. The resulting astrophysical reaction rate is 20%–30% larger than the currently recommended rate at s-process temperatures, which should lead to better agreement between stellar models and meteoric isotopic anomalies. The neutron-capture cross section at low energies is substantially larger than recommended in the latest evaluation, which is problematical for criticality benchmarks. The average resonance spacing as a function of spin and parity is significantly different from current models. The total-radiation-width distributions are significantly broader than predicted by theory and show significant departures from the expected Gaussian shapes.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Wallula Basalt Pilot Demonstration Project: Post-injection Results and Conclusions

Deep underground geologic formations are emerging as a reasonable option for long-term storage of CO 2 , including large continental flood basalt formations. At the GHGT-11 and GHGT-12 conferences, progress was reported on the initial phases for Wallula Basalt Pilot demonstration test (located in Eastern Washington state), where nearly 1,000 metric tons of CO 2 were injected over a 3-week period during July/August 2013. The target CO 2 injection intervals were two permeable basalt interflow reservoir zones with a combined thickness of ~20 m that occur within a layered basalt sequence between a depth of 830-890 m below ground surface. During the two-year post-injection period, downhole fluid samples were periodically collected during this post-injection monitoring phase, coupled with limited wireline borehole logging surveys that provided indirect evidence of on-going chemical geochemical reactions/alterations and CO 2 disposition. A final detailed post-closure field characterization program that included downhole fluid sampling, and performance of hydrologic tests and wireline geophysical surveys. Included as part of the final wireline characterization activities was the retrieval of side-wall cores from within the targeted injection zones. These cores were examined for evidence of in-situ mineral carbonization. Visual observations of the core material identified small globular nodules, translucent to yellow in color, residing within vugs and small cavities of the recovered basalt side-wall cores, which were not evident in pre-injection side-wall cores obtained from the native basalt formation. Characterization by x-ray diffraction identified these nodular precipitates as ankerite, a commonly occurring iron and calcium rich carbonate. Isotopic characterization (δ 13 C, δ 18 O) conducted on the ankerite nodules indicate a distinct isotopic signature that is closely aligned with that of the injected CO 2 . Both the secondary mineral nodules and injected CO 2 are measurably different from the isotopic content of basalt, injection zone groundwater and for naturally occurring calcite. Final post-injection wireline geophysical logging results also indicate the presence of free-phase CO 2 at the top of the two injection interflow zones, with no vertical migration of CO 2 above the injection horizons. Furthermore, these findings are significant and demonstrate the feasibility of sequestering CO 2 in a basalt formation.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Current status of tritium calorimetry at TLK

Inside a tritium facility, calorimetry is an important analytical method as it is the only reference method for accountancy (it is based on the measurement of the heat generated by the radioactive decay). Presently, at Tritium Laboratory Karlsruhe (TLK), 4 calorimeters are in operation, one of isothermal type and three of inertial guidance control type (IGC). The volume of the calorimeters varies between 0.5 and 20.6 liters. About two years ago we started an extensive work to improve our calorimeters with regard to reliability and precision. We were forced to upgrade 3 of our 4 calorimeters due to the outdated interfaces and software. This work involved creating new LabView programs driving the devices, re-tuning control loops and replacing obsolete hardware components. In this paper we give a review on the current performance of our calorimeters, comparing it to recently available devices from the market and in the literature. We also show some ideas for a next generation calorimeter based on experiences with our IGC calorimeters and other devices reported in the literature. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

Post-Irradiation Examination of 237Np Targets for 238Pu Production

Oak Ridge National Laboratory is recovering the US 238 Pu production capability and the first step in the process has been to evaluate the performance of a 237 Np target cermet pellet encased in an aluminum clad. The process proceeded in 3 steps; the first step was to irradiate capsules of single pellets composed of NpO 2 and aluminum power to examine their shrinkage and gas release. These pellets were formed by compressing sintered NpO 2 and aluminum powder in a die at high pressure followed by sintering in a vacuum furnace. Three temperatures were chosen for sintering the solution precipitated NpO 2 power used for pellet fabrication. The second step was to irradiate partial targets composed of 8 pellets in a semi-prototypical arrangement at the two best performing sintering temperatures to determine which temperature gave a pellet that performed the best under the actual planned irradiation conditions. The third step was to irradiate ~50 pellets in an actual target configuration at design irradiation conditions to assess pellet shrinkage and gas release, target heat transfer, and dimensional stability. The higher sintering temperature appeared to offer the best performance after one cycle of irradiation by having the least shrinkage, thus keeping the heat transfer gap between the pellets and clad small minimizing the pellet operating temperature. The final result of the testing was a target that can meet the initial production goals, satisfy the reactor safety requirements, and can be fabricated in production quantities. The current focus of the program is to verify that the target can be remotely dissembled, the pellets dissolved, and the 238 Pu recovered. Tests are being conducted to examine these concerns and to compare results to code predictions. Once the performance of the full length targets has been quantified, the pellet 237 Np loading will be revisited to determine if it can be increased to increase 238 Pu production.

07 ISOTOPE AND RADIATION SOURCES↗