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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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108 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Solid-liquid Separation Testing for the Remediation of Hanford Waste Treatment Plant Low Activity Waste Melter Off-Gas Condensate

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during Direct Feed LAW operations is to transfer it to the Effluent Management Facility (EMF), where it will be concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. However, this stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the quantity of glass waste produced by allowing further increases in waste loading, resulting in a shortened LAW vitrification mission duration. The concept being tested here involves selectively removing the 99 Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed as a different waste form and avoid reprocessing through vitrification.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Development of a patient-specific dosimetry estimation system in nuclear medicine examination

The purpose of this study is to develop a patient-specific dosimetry estimation system in nuclear medicine examination using a SimSET-based Monte Carlo code. We added a dose deposition routine to store the deposited energy of the photons during their flights in SimSET and developed a user-friendly interface for reading PET and CT images. Dose calculated on ORNL phantom was used to validate the accuracy of this system. The S values for {sup 99m}Tc, {sup 18}F and {sup 131}I obtained by the system were compared to those from the MCNP4C code and OLINDA. The ratios of S values computed by this system to those obtained with OLINDA for various organs were ranged from 0.93 to 1.18, which are comparable to that obtained from MCNP4C code (0.94 to 1.20). The average ratios of S value were 0.99{+-}0.04, 1.03{+-}0.05, and 1.00{+-}0.07 for isotopes {sup 131}I, {sup 18}F, and {sup 99m}Tc, respectively. The simulation time of SimSET was two times faster than MCNP4C's for various isotopes. A 3D dose calculation was also performed on a patient data set with PET/CT examination using this system. Results from the patient data showed that the estimated S values using this system differed slightly from those of OLINDA for ORNL phantom. In conclusion, this system can generate patient-specific dose distribution and display the isodose curves on top of the anatomic structure through a friendly graphic user interface. It may also provide a useful tool to establish an appropriate dose-reduction strategy to patients in nuclear medicine environments. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

SU-E-CAMPUS-I-05: Internal Dosimetric Calculations for Several Imaging Radiopharmaceuticals in Preclinical Studies and Quantitative Assessment of the Mouse Size Impact On Them. Realistic Monte Carlo Simulations Based On the 4D-MOBY Model

Purpose: Commonly used radiopharmaceuticals were tested to define the most important dosimetric factors in preclinical studies. Dosimetric calculations were applied in two different whole-body mouse models, with varying organ size, so as to determine their impact on absorbed doses and S-values. Organ mass influence was evaluated with computational models and Monte Carlo(MC) simulations. Methods: MC simulations were executed on GATE to determine dose distribution in the 4D digital MOBY mouse phantom. Two mouse models, 28 and 34 g respectively, were constructed based on realistic preclinical exams to calculate the absorbed doses and S-values of five commonly used radionuclides in SPECT/PET studies (18F, 68Ga, 177Lu, 111In and 99mTc).Radionuclide biodistributions were obtained from literature. Realistic statistics (uncertainty lower than 4.5%) were acquired using the standard physical model in Geant4. Comparisons of the dosimetric calculations on the two different phantoms for each radiopharmaceutical are presented. Results: Dose per organ in mGy was calculated for all radiopharmaceuticals. The two models introduced a difference of 0.69% in their brain masses, while the largest differences were observed in the marrow 18.98% and in the thyroid 18.65% masses.Furthermore, S-values of the most important target-organs were calculated for each isotope. Source-organ was selected to be the whole mouse body.Differences on the S-factors were observed in the 6.0–30.0% range. Tables with all the calculations as reference dosimetric data were developed. Conclusion: Accurate dose per organ and the most appropriate S-values are derived for specific preclinical studies. The impact of the mouse model size is rather high (up to 30% for a 17.65% difference in the total mass), and thus accurate definition of the organ mass is a crucial parameter for self-absorbed S values calculation.Our goal is to extent the study for accurate estimations in small animal imaging, whereas it is known that there is a large variety in the anatomy of the organs.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

Solid Secondary Waste Form Testing Activities for the Integrated Disposal Facility Performance Assessment Support - 17567

Solid secondary waste (SSW) that will be generated at the Waste Treatment and Immobilization Plant (WTP) and from other activities at the Hanford site is destined for disposal at the Integrated Disposal Facility (IDF). Construction of the IDF was undertaken based on a Performance Assessment (PA) performed in 2001. In 2016, a SSW data package was prepared to support the development of an updated PA (2017 IDF PA). There are many inputs required to support the IDF PA. Data needs were identified based on experience from the previous PAs and from the development of conceptual models and modeling approaches to be applied for the 2017 PA. Interactions with the IDF PA team and the inventory data package developer helped to focus detailed efforts on a few key waste streams, contaminants of concern, and disposal configurations that are expected to be the primary contributors to the dose resulting from SSW disposed at the IDF. The key waste streams include: High Efficiency Particulate Air (HEPA) filters, ion exchange resins, granular activated carbon, and Ag-mordenite. The key parameters addressed in the 2016 data package include: saturated hydraulic conductivity, moisture characteristic curves, effective diffusion coefficient, distribution coefficient and porosity. The IDF PA team also identified a set of key contaminants that are expected to be the primary concern for the PA calculations (Tc-99, I-129, Cs-137, Sr-90, uranium isotopes (and total uranium), chromium, mercury and nitrate). These contaminants were the focus of the SSW data package. Specific formulations had not been identified for cementitious materials that will be used to encapsulate or stabilize SSW, and no IDF-specific experiments had been conducted to obtain the data for the PA. Thus, for the 2016 SSW data package, recommended property values were provided for a range of representative candidate materials based on a review of existing literature. However, starting in FY17, IDF-specific experiments will be performed for Hanford SSW expected to contribute most significantly to the dose for the IDF. This paper will summarize the information obtained for the 2016 data package and the plans for further testing to refine key inputs for the IDF PA. (authors)

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

Radiogenic p-isotopes from type Ia supernova, nuclear physics uncertainties, and galactic chemical evolution compared with values in primitive meteorites

The nucleosynthesis of proton-rich isotopes is calculated for multi-dimensional Chandrasekhar-mass models of Type Ia supernovae (SNe Ia) with different metallicities. The predicted abundances of the short-lived radioactive isotopes {sup 92}Nb, {sup 97,} {sup 98}Tc, and {sup 146}Sm are given in this framework. The abundance seeds are obtained by calculating s-process nucleosynthesis in the material accreted onto a carbon-oxygen white dwarf from a binary companion. A fine grid of s-seeds at different metallicities and {sup 13}C-pocket efficiencies is considered. A galactic chemical evolution model is used to predict the contribution of SN Ia to the solar system p-nuclei composition measured in meteorites. Nuclear physics uncertainties are critical to determine the role of SNe Ia in the production of {sup 92}Nb and {sup 146}Sm. We find that, if standard Chandrasekhar-mass SNe Ia are at least 50% of all SN Ia, they are strong candidates for reproducing the radiogenic p-process signature observed in meteorites.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2020

Sorption of Anionic Iodine and Molybdenum to Multi-sorbing Barrier (MSB) Materials - 20030

A multi-sorbing barrier (MSB) material containing sodium bentonite (NaB), organoclay, and anion resin is being evaluated for containment of anionic radionuclides. Batch tests on the MSB were conducted using two non-radioactive analog isotopes ({sup 127}I, {sup 95}Mo) as surrogates for {sup 129}I and {sup 99}Tc due to their similar valence and molecular weight relative to the radionuclides. Tests were conducted on the organoclay, anion resin, and NaB individually and on NaB-resin blends with resin contents of 1% and 5% by mass. The anion resin exhibits very strong sorption of {sup 127}I and {sup 95}Mo relative to the other materials. The organoclay exhibited modest sorption of {sup 95}Mo and {sup 127}I, and NaB little sorption. The sorption isotherms follow a linear model for NaB, a Freundlich model for the organoclay, and a Langmuir model for the anion resin and NaB-resin mixtures. Combining resin with NaB significantly enhances sorption relative to NaB alone, with a larger resin fraction yielding greater sorption. (authors)

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

MO-G-17A-04: Internal Dosimetric Calculations for Pediatric Nuclear Imaging Applications, Using Monte Carlo Simulations and High-Resolution Pediatric Computational Models

Purpose: Our purpose is to evaluate the administered absorbed dose in pediatric, nuclear imaging studies. Monte Carlo simulations with the incorporation of pediatric computational models can serve as reference for the accurate determination of absorbed dose. The procedure of the calculated dosimetric factors is described, while a dataset of reference doses is created. Methods: Realistic simulations were executed using the GATE toolkit and a series of pediatric computational models, developed by the “IT'IS Foundation”. The series of the phantoms used in our work includes 6 models in the range of 5–14 years old (3 boys and 3 girls). Pre-processing techniques were applied to the images, to incorporate the phantoms in GATE simulations. The resolution of the phantoms was set to 2 mm3. The most important organ densities were simulated according to the GATE “Materials Database”. Several used radiopharmaceuticals in SPECT and PET applications are being tested, following the EANM pediatric dosage protocol. The biodistributions of the several isotopes used as activity maps in the simulations, were derived by the literature. Results: Initial results of absorbed dose per organ (mGy) are presented in a 5 years old girl from the whole body exposure to 99mTc - SestaMIBI, 30 minutes after administration. Heart, kidney, liver, ovary, pancreas and brain are the most critical organs, in which the S-factors are calculated. The statistical uncertainty in the simulation procedure was kept lower than 5%. The Sfactors for each target organ are calculated in Gy/(MBq*sec) with highest dose being absorbed in kidneys and pancreas (9.29*10{sup 10} and 0.15*10{sup 10} respectively). Conclusion: An approach for the accurate dosimetry on pediatric models is presented, creating a reference dosage dataset for several radionuclides in children computational models with the advantages of MC techniques. Our study is ongoing, extending our investigation to other reference models and evaluating the results with clinical estimated doses.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗
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 OSTI2017

Precipitation of Epsilon Particles in Metal Doped Ceria Irradiated with Energetic Ions

Metallic particles precipitated from fission products in oxide fuel matrix (e.g., UO{sub 2}) have a great impact on the thermal and mechanical properties of the fuel. A burn-up of 40 MWd/kg U results in the conversion of 4% of the uranium to approximately 3% fission products and 1% transuranium elements. Epsilon phase is the major metallic phase of the precipitates with a hexagonal-closed packed (HCP) crystal structure. It can exist in a variety of metallic forms in UO{sub 2}, but typically, it consists of five metals: Mo, Pd, Rh, Ru and Tc. Epsilon particles can form over a wide range of stoichiometries in the five metal system. Simulated compositions with Re as a surrogate for Tc have been investigated as a waste form for {sup 99}Tc, an isotope with a long half-life of 2.13 10{sup 5} years and a high solubility under oxidizing conditions in addition to Mo in the corrosion process. There is a pressing need to understand the factors influencing the formation and growth of the metallic phase in fuel. Efforts are currently underway to perform careful neutron irradiation of nuclear fuel under controlled conditions in a test reactor, recover metallic phases that may be present, and characterize the resulting structure and chemistry. Due to the long duration, high cost and radiological concerns associated with neutron irradiation and sample characterizations, ion irradiation studies of a surrogate material with a full control of the processing parameters can emulate different stages of the particle precipitation, which can add a tremendous value to reactor-based studies. This presentation reports our recent ion irradiation results on the metallic particle precipitation under various irradiation and thermal annealing conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

A dose point kernel database using GATE Monte Carlo simulation toolkit for nuclear medicine applications: Comparison with other Monte Carlo codes

Purpose: GATE is a Monte Carlo simulation toolkit based on the Geant4 package, widely used for many medical physics applications, including SPECT and PET image simulation and more recently CT image simulation and patient dosimetry. The purpose of the current study was to calculate dose point kernels (DPKs) using GATE, compare them against reference data, and finally produce a complete dataset of the total DPKs for the most commonly used radionuclides in nuclear medicine. Methods: Patient-specific absorbed dose calculations can be carried out using Monte Carlo simulations. The latest version of GATE extends its applications to Radiotherapy and Dosimetry. Comparison of the proposed method for the generation of DPKs was performed for (a) monoenergetic electron sources, with energies ranging from 10 keV to 10 MeV, (b) beta emitting isotopes, e.g., {sup 177}Lu, {sup 90}Y, and {sup 32}P, and (c) gamma emitting isotopes, e.g., {sup 111}In, {sup 131}I, {sup 125}I, and {sup 99m}Tc. Point isotropic sources were simulated at the center of a sphere phantom, and the absorbed dose was stored in concentric spherical shells around the source. Evaluation was performed with already published studies for different Monte Carlo codes namely MCNP, EGS, FLUKA, ETRAN, GEPTS, and PENELOPE. A complete dataset of total DPKs was generated for water (equivalent to soft tissue), bone, and lung. This dataset takes into account all the major components of radiation interactions for the selected isotopes, including the absorbed dose from emitted electrons, photons, and all secondary particles generated from the electromagnetic interactions. Results: GATE comparison provided reliable results in all cases (monoenergetic electrons, beta emitting isotopes, and photon emitting isotopes). The observed differences between GATE and other codes are less than 10% and comparable to the discrepancies observed among other packages. The produced DPKs are in very good agreement with the already published data, which allowed us to produce a unique DPKs dataset using GATE. The dataset contains the total DPKs for {sup 67}Ga, {sup 68}Ga, {sup 90}Y, {sup 99m}Tc, {sup 111}In, {sup 123}I, {sup 124}I, {sup 125}I, {sup 131}I, {sup 153}Sm, {sup 177}Lu {sup 186}Re, and {sup 188}Re generated in water, bone, and lung. Conclusions: In this study, the authors have checked GATE's reliability for absorbed dose calculation when transporting different kind of particles, which indicates its robustness for dosimetry applications. A novel dataset of DPKs is provided, which can be applied in patient-specific dosimetry using analytical point kernel convolution algorithms.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2014

Sci—Thur PM: Imaging — 04: An iterative triple energy window (TEW) approach to cross talk correction in quantitative small animal Tc99m and In111 SPECT

Dual isotope SPECT allows simultaneous measurement of two different tracers in vivo. With In111 (emission energies of 171keV and 245keV) and Tc99m (140keV), quantification of Tc99m is degraded by cross talk from the In111 photons that scatter and are detected at an energy corresponding to Tc99m. TEW uses counts recorded in two narrow windows surrounding the Tc99m primary window to estimate scatter. Iterative TEW corrects for the bias introduced into the TEW estimate resulting from un-scattered counts detected in the scatter windows. The contamination in the scatter windows is iteratively estimated and subtracted as a fraction of the scatter-corrected primary window counts. The iterative TEW approach was validated with a small-animal SPECT/CT camera using a 2.5mL plastic container holding thoroughly mixed Tc99m/In111 activity fractions of 0.15, 0.28, 0.52, 0.99, 2.47 and 6.90. Dose calibrator measurements were the gold standard. Uncorrected for scatter, the Tc99m activity was over-estimated by as much as 80%. Unmodified TEW underestimated the Tc99m activity by 13%. With iterative TEW corrections applied in projection space, the Tc99m activity was estimated within 5% of truth across all activity fractions above 0.15. This is an improvement over the non-iterative TEW, which could not sufficiently correct for scatter in the 0.15 and 0.28 phantoms.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2013

Nuclear Data Sheets for A = 89

The evaluated experimental data are presented for 14 known nuclides of mass 89 (Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh). New data on high–spin excitations are available for {sup 89}Kr, {sup 89}Rb, {sup 89}Sr, {sup 89}Tc and {sup 89}Ru, including a superdeformed structure in {sup 89}Tc. Fluorescence studies in {sup 89}Y have revealed a large number of new levels in 7–11 MeV range. New two–neutron transfer data are available for {sup 89}Zr. New direct and precise measurement of atomic masses of {sup 89}Se, {sup 89}Br, 89{sup K}r, {sup 89}Rb, {sup 89}Mo and {sup 89}Tc have greatly improved the Q value landscape in this mass region. This work supersedes previous A=89 evaluation published in 1998Si31, and also the older ones 1989Si20, 1975Ko21, and M.W. Johns et al., Nuclear Data Tables A 8, 373 (1970). In spite of extensive experimental work on the isobaric nuclei of this mass chain several deficiencies remain. The identification of {sup 89}Rh isotope remains unconfirmed. The half–lives of ground states of {sup 89}Ge, {sup 89}As and {sup 89}Rh have not been measured, only the lower limits are estimated from time–of–flight in a reaction and experimental arrangement, where produced and identified. The decay schemes of {sup 89}Ge, {sup 89}As, {sup 89}Se, {sup 89}Ru and {sup 89}Rh are not known, and those for {sup 89}Tc and {sup 89}Mo are incomplete. High–lying (neutron unbound) levels in {sup 89}Kr, expected to be populated in the decay of {sup 89}Br, have not been investigated. The spins and parities for levels in {sup 89}Kr are largely unassigned, while for {sup 89}Tc and {sup 89}Ru, only few low–spin excitations are established. For {sup 89}Rb, little information is available for high–spin structures. Detailed gamma–ray data are available for {sup 88}Sr(n,γ){sup 89}Sr reaction, but most of these γ rays remain unassigned in a level scheme. Excited state data are nonexistent for {sup 89}Ge, {sup 89}As, {sup 89}Se, {sup 89}Br and {sup 89}Rh. The {sup 89}Y and {sup 89}Zr are the most extensively studied nuclei in A=89.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Flow Measurements in Pressurized Helium Gas Cooling Channels

Production of metastable Technetium-99 (Tc-99m) is vital to the medical imaging community. It is a radioactive tracer that emits gamma rays as it decays and it is these gamma rays that are then detected through imaging. Tc-99m is extracted from the decay of Molybdenum-99 (Mo-99) which has a half-life of about 2-3 days. One method of generating Mo-99 is using accelerator based technology, through the irradiation of Mo-100 using an electron beam. This is part of the NNSA’s mission to produce Mo-99 commercially, within the US, without the use of highly enriched uranium (HEU) in support of nonproliferation and global security. Los Alamos National Laboratory (LANL) is working with NorthStar medical Radioisotopes (NMR) on their efforts to produce Mo-99 from the irradiation of Mo-100 targets. The NMR target consists of an Inconel window that allows the electron beam to penetrate and irradiate a stack of Mo-100 discs. The irradiation process generates large amounts of heat and requires extensive cooling of the Mo-100 targets and target housing using pressurized helium gas. Before beam on target, the design of the target stack and target housing is tested at LANL using laser based diagnostics experiments. We quantify the velocity field of the helium gas around the target housing to estimate whether the cooling provided is sufficient for the design and to validate computational fluid dynamics (CFD) models. Fiber optic imaging is also developed for use during cold testing (without beam on target) as well as during irradiation, to obtain real time information on disc performance.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Experiment on the treatment of waste extraction solvent from the molybdenum-99 process

In the Mo-99 (Molybdenum-99) isotope extraction test process for radiopharmaceutical applications, organic solvent is used to extract Mo-99 from an irradiated UO{sub 2} dissolution. The extraction solvent was stored when the test work was stopped. A total of about 120 liters of waste solvent was stored at INER (Institute of Nuclear Energy Research, Taiwan). The extraction solvent consisted of 5% di-(2-ethylhexyl)-phosphoric acid (D2EHPA) and kerosene. The radionuclides found in the waste solvent include Cs-137, Am-241, Tc-99, and Sr-90, which give off gross alpha and beta radioactivity of 1898 and 471 Bq/ml, respectively. This study aims to remove radionuclides from the waste solvent using sodium carbonate and sodium hydroxide solutions in different concentrations. After mixing the waste solvent with the alkaline solution followed by settling, a third phase other than organic and aqueous phase appeared which is expected due to the saponification reaction. The experimental results showed that increasing the number of washing and the alkaline solution concentration could enhance the radionuclides removal rate. An optimal removal method was proposed using 2 M Na{sub 2}CO{sub 3} solution twice followed by 1 M NaOH solution one time for the third phase generated early in the mixing stages. The remaining gross alpha and beta radioactivity of the treated organic solvent was 2 and 3 Bq/ml, respectively. The treated solvent could be stabilized by ashing at 500 deg. C and then immobilized. The alkaline solution would be neutralized by hydrochloric or nitric acid and then treated using a variety of adsorbents or bone char via adsorption to remove nuclides to meet the wastewater discharge limitation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Synthesis, Structure Elucidation, and Redox Properties of [superscript 99]Tc Complexes of Lacunary Wells-Dawson Polyoxometalates: Insights into Molecular [superscript 99]Tc-Metal Oxide Interactions

The isotope {sup 99}Tc ({beta}{sub max}, 293.7; half-life, 2.1 x 10{sup 5} years) is an abundant product of uranium-235 fission in nuclear reactors and is present throughout the radioactive waste stored in underground tanks at the Hanford and Savannah River sites. Understanding and controlling the extensive redox chemistry of {sup 99}Tc is important in identifying tunable strategies to separate {sup 99}Tc from spent fuel and from waste tanks and, once separated, to identify and develop an appropriately stable waste form for {sup 99}Tc. Polyoxometalates (POMs), nanometer-sized models for metal oxide solid-state materials, are used in this study to provide a molecular level understanding of the speciation and redox chemistry of incorporated {sup 99}Tc. In this study, {sup 99}Tc complexes of the ({alpha}{sub 2}-P{sub 2}W{sub 17}O{sub 61}){sup 10-} and ({alpha}{sub 1}-P{sub 2}W{sub 17}O{sub 61}){sup 10-} isomers were prepared. Ethylene glycol was used as a 'transfer ligand' to minimize the formation of TcO{sub 2} {center_dot} xH{sub 2}O. The solution structures, formulations, and purity of TcVO({alpha}{sub 1}/{alpha}{sub 2}-P{sub 2}W{sub 17}O{sub 61}){sup 7-} were determined by multinuclear NMR. X-ray absorption spectroscopy of the complexes is in agreement with the formulation and structures determined from {sup 31}P and {sup 183}W NMR. Preliminary electrochemistry results are consistent with the EXAFS results, showing a facile reduction of the TcVO({alpha}{sub 1}-P{sub 2}W{sub 17}O{sub 61}){sup 7-} species compared to the TcVO({alpha}{sub 2}-P{sub 2}W{sub 17}O{sub 61}){sup 7-} analog. The {alpha}{sub 1} defect is unique in that a basic oxygen atom is positioned toward the {alpha}{sub 1} site, and the Tc{sup V}O center appears to form a dative metal-metal bond with a framework W site. These attributes may lead to the assistance of protonation events that facilitate reduction. Electrochemistry comparison shows that the ReV analogs are about 200 mV more difficult to reduce in accordance with periodic trends.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Advancing Performance Assessment for Disposal of Depleted Uranium at Clive Utah - 12493

A Performance Assessment (PA) for disposal of depleted uranium (DU) waste has recently been completed for a potential disposal facility at Clive in northwestern Utah. For the purposes of this PA, 'DU waste' includes uranium oxides of all naturally-occurring isotopes, though depleted in U-235, varying quantities of other radionuclides introduced to the uranium enrichment process in the form of used nuclear reactor fuel (reactor returns), and decay products of all of these radionuclides. The PA will be used by the State of Utah to inform an approval decision for disposal of DU waste at the facility, and will be available to federal regulators as they revisit rulemaking for the disposal of DU. The specific performance objectives of the Clive DU PA relate to annual individual radiation dose within a 10,000-year performance period, groundwater concentrations of specific radionuclides within a 500-year compliance period, and site stability in the longer term. Fate and transport processes that underlie the PA model include radioactive decay and ingrowth, diffusion in gaseous and water phases, water advection in unsaturated and saturated zones, transport caused by plant and animal activity, cover naturalization, natural and anthropogenic erosion, and air dispersion. Fate and transport models were used to support the dose assessment and the evaluation of groundwater concentrations. Exposure assessment was based on site-specific scenarios, since the traditional human exposure scenarios suggested by DOE and NRC guidance are unrealistic for this site. Because the U-238 in DU waste reaches peak radioactivity (secular equilibrium) after 2 million years (My) following its separation, the PA must also evaluate the impact of climate change cycles, including the return of pluvial lakes such as Lake Bonneville. The first draft of the PA has been submitted to the State of Utah for review. The results of this preliminary analysis indicate that doses are very low for the site-specific receptors for the 10,000-year compliance period. This is primarily because DU waste is not highly radioactive within this time frame, the DU waste is assumed to be buried beneath zones exposed by erosion, groundwater concentrations of DU waste constituents do not exceed groundwater protection limits with in the 500-year compliance period, and the first deep lake occurrence will disperse DU waste across a large area, and will ultimately be covered by lake-derived sediment. A probabilistic PA model was constructed that considered DU waste and decay product doses to site-specific receptors for a 10,000-yr performance period, as well as deep-time effects. The quantitative results are summarized in Table VII. Doses (as TEDE) are always less than 5 mSv in a year, and doses to the offsite receptors are always much less than 0.25 mSv in a year. Groundwater concentrations of Tc-99 are always less than its GWPL except when the Tc-99 contaminated waste is disposed below grade. Even in this case, the median groundwater concentration is only 4.18 Bq/L (113 pCi/L), which is more than one order of magnitude less than the GWPL for Tc-99. The results overall suggest that there are disposal configurations that can be used to dispose of the proposed quantities of DU waste that are adequately protective of human health. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗