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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Technetium Chemistry in the Fuel Cycle: Combining Basic and Applied Studies

Technetium is intimately linked with nuclear reactions. The ultraminute natural levels in the environment are due to the spontaneous fission of uranium isotopes. The discovery of technetium was born from accelerator reactions, and its use and presence in the modern world are directly due to nuclear reactors. While occupying a central location in the periodic table, the chemistry of technetium is poorly explored, especially when compared to its neighboring elements, i.e., molybdenum, ruthenium, and rhenium. This state of affairs, which is tied to the small number of laboratories equipped to work with the long-lived (99)Tc isotope, provides a remarkable opportunity to combine basic studies with applications for the nuclear fuel cycle. An example is given through examination of the technetium halide compounds. Binary metal halides represent some of the most fundamental of inorganic compounds. The synthesis of new technetium halides demonstrates trends with structure, coordination number, and speciation that can be utilized in the nuclear fuel cycle. Examples are provided for technetium-zirconium alloys as waste forms and the formation of reduced technetium species in separations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Characterization of Tc-99 Speciation and Chemical Composition of 241-AN-102 Tank Waste Supernatant - 19201

Technetium (Tc) in its isotopic form {sup 99}Tc is generated in high yields from the fission of {sup 235}U and {sup 239}Pu. A significant inventory of {sup 99}Tc remains at the U.S. Department of Energy's Hanford Site, and it is one of the most difficult contaminants to address. In the strongly alkaline solutions that are typical of Hanford tank waste, Tc exists as both pertechnetate (TcO{sub 4}{sup -} in oxidation state VII) and in reduced forms (oxidation state <VII) collectively known as non-pertechnetate species. Pertechnetate is a well-characterized anionic Tc species that can be removed from low-activity waste by anion exchange or other methods. There is no definitive information on the origin or a comprehensive description of the non-pertechnetate species in Hanford tanks. Designing strategies for effective management of Tc, including separation and immobilization, requires an understanding of the molecular structure of the non-pertechnetate species and their identification in actual tank waste samples. Developing this advanced understanding would facilitate development of new technologies for treating dissimilar Tc species. Toward this objective, a spectroscopic library of the Tc(I) [fac-Tc(CO){sub 3}]{sup +} and Tc(IV, VI, VII) compounds was generated using a range of techniques and applied to the characterization of actual 241-AN-102 tank waste supernatant. A sample of the 241-AN-102 tank waste (collected in 2013) was processed to remove {sup 137}Cs by spherical resorcinol-formaldehyde (sRF) ion exchange resin after the sodium concentration was adjusted to 5.6 M. The Cs-loaded sRF column was eluted with 0.45 M HNO{sub 3}. As-received AN-102, Cs-depleted AN-102 effluent, and sRF eluate fractions were comprehensively characterized for chemical composition and speciation of Tc using {sup 99}Tc nuclear magnetic resonance (NMR) spectroscopy and X-ray absorption spectroscopy. The results of these analyses demonstrated for the first time that non-pertechnetate Tc present in the 241-AN-102 tank waste is composed of several low-valent Tc species, including the Tc(I) [fac-Tc(CO){sub 3}L{sub 3}]{sup +}, Tc(IV), and possibly Tc(VI) compounds. This is the second experimental observation of the [fac-Tc(CO){sub 3}L{sub 3}]{sup +} species in the Hanford tank waste that highlights their importance for waste processing. In addition, the chemical composition of the 241-AN-102 tank waste was characterized with the focus on small chelators. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Vadose and Saturated Zone Flow and Transport Calculations for the Active Trenches of the Low-Level Burial Grounds, Hanford Site, Washington

The purpose of the fate and transport modeling described in this environmental calculation file (ECF) is to evaluate the impacts to groundwater associated with waste disposal operations at Mixed Waste Low-Level Burial Ground (LLBG) Trenches 31 and 34 to satisfy requirements in DOE O 435.1, Radioactive Waste Management. The model integrates the flow and transport in the vadose zone beneath the active trenches with the saturated zone downgradient of the trenches to predict the radionuclide concentration at the point of assessment (POA). DOE M 435.1-1, Radioactive Waste Management Manual, defines the POA as the point of highest projected dose or concentration beyond a 100 m (328 ft) buffer zone surrounding the disposed waste. The modeling is conducted in accordance with the DOE G 435.1, Implementation Guide for Use with DOE M 435.1-1, performance assessment (PA) guidelines. The modeling involves evaluation of the groundwater concentrations and radionuclide arrival times during the 1,000-year compliance and 10,000-year sensitivity-uncertainty periods per DOE O 435.1 and DOE M 435.1. This analysis does not consider radionuclide release during facility operations, only the post-closure impacts of the radionuclides to the environment. The evaluation of potential radiological dose to groundwater receptors caused by releases from a closed facility containing radioactive waste typically includes the following: (1) Release of radionuclides from that facility (2) Transport of those radionuclides through the environment, and (3) Exposure to humans to environmental concentration levels of those radionuclides The fate and transport three-dimensional (3D) model analysis involves the post-closure impacts to the environment of the technetium-99, iodone-129, and uranium (all isotopes in the waste). The residual inventory estimates include several radionuclides, but technetium-99 is typically responsible for almost all of the beta-gamma dose equivalent associated with groundwater (water resources) protection per 40 CFR 141, “National Primary Drinking Water Regulations” (e.g., see the results in WCH-520, Performance Assessment of Environmental Restoration Disposal Facility, Hanford Site, Washington; hereinafter referred to as the ERDF PA), and iodone-129 can also be a significant dose contributor for some waste (e.g., RPP-RPT-59958, Performance Assessment for the Integrated Disposal Facility, Hanford Site, Washington; hereinafter referred to as the IDF PA). Uranium does not typically factor significantly into the impacts to groundwater, even during the 10,000-year sensitivity-uncertainty period, but always remains of interest as a contaminant. This ECF does not address vadose and saturated zone modeling for Trench 94 of the 200 East Area LLBG. Current information confirms the validity of the low corrosion rate of the naval reactor plant carbon steel (HY-80), and the even lower corrosion rate of the nickel-iron-chromium alloy reactor vessel (Inconel Alloy 600) presented in DOE/EIS-0259, Final Environmental Impact Statement on the Disposal of Decommissioned, Defueled Cruiser, Ohio Class, and Los Angeles Class Naval Reactor Plants. Based on these low corrosion rates, the time to breach the reactor vessel to allow release of radionuclides from the activated metal of the reactor vessel internal structure is at least 10,000 years. This time to breach precludes the need to evaluate the vadose and saturated zone transport of contaminants released from the reactor compartment disposal packages in the 200 East Area LLBGs PA (CP-63826, Waste Release Model Package Report for the Active Trenches of the Low-Level Burial Grounds, Hanford Site, Washington).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Moly 99 Reactor Case Study

Molybdenum-99 (Moly 99) is a critical raw material for Technetium (Tc) 99m, a radioactive isotope most widely used in nuclear medicine procedures. Moly 99 has a short half-life of about six hours, which means it cannot be stockpiled. When Moly 99 decays, it turns into Tc 99m, which has a half-life of 214,000 years. The photon energy emitted from the decay of Moly 99 is used in a variety of nuclear imaging technologies such as gamma cameras. Furthermore, radiopharmaceutical manufacturers use the photon energy emitted from the decay of Moly 99 to produce generators for hospitals, clinics, and radiopharmacies. Once Moly 99 decays to Tc 99m, it is used to make individual patient doses for a variety of diagnostic imaging procedures. The Moly 99 Reactor Design is a conceptual blueprint for Moly 99 production that does not use weapon-grade uranium; instead, the reactor has a target core of low-enriched uranium. The reactor design is small, reaching a foot-and-a-half in height and diameter and consumes less than two megawatts of power. The reactor's only purpose is for medical isotope production and with every fission, Moly 99 is produced.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2016

Tin(II)apatite: Synthesis, Characterization, and Challenge with Pertechnetate - 16194

We will report on the synthesis of stannous apatite and the subsequent reductive capacity attained, using a cerium (IV) titration method. Also reported will be the distribution coefficients derived using a Hanford tank supernatant simulant of 7.8 M sodium with chrome (VI) as a contaminant of concern, as a potential non-isotopic oxy-anion and technetium surrogate for reductive capacity determinations. Pertechnetate will be subjected to reduction using hydroxylamine, glucose and threose. The purpose is to take the first step in laboratory analyses to determine the amount of non-pertechnetate species in Hanford tank waste. A proposed sequestration mechanism will be presented as to why the reacted tin(II)apatite does not release technetium once reacted. SEM and EDS spectra will be presented on the crystalline apatite. Areas of incorporation of the tin(II)apatite will be presented for the Hanford tank waste process. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

The Thermophysical Properties of TcO2

Technetium-99 is a highly radioactive isotope with a long half-life that is common in nuclear waste. It volatizes at a low temperature, which poses a significant challenge to the clean-up and containment processes. Due to difficulties in purifying technetium compounds, their thermophysical properties have not been measured or calculated. Here, first principle methods are used along with the quasi quasi-harmonic harmonic approximation to compute the Debye temperature, volumetric thermal expansion coefficient, bulk modulus, and heat capacity of rutile TcO2 for temperatures ranging from 0 to 1500 K and applied pressures ranging from 0 to 255 GPa. The computed atomic structures agree well with the results from diffraction measurements. The computed thermophysical properties are in the neighborhood of other rutile metal oxides and, in particular, are within approximately 10–13% of rutile ReO2, which is frequently used as a substitute for TcO2 in experimental studies.

Zhong, Hong↗
DOE OSTI2020

Design Optimization of A generic Fissile Solution for Mo99 production using Electron Beam-based Neutron Generator using MCNP+CFD

Molybdenum-99 ( 99 Mo) is a medical isotope used in 80% of all medical imaging procedures today. However, 99 Mo is not directly administered to patients for imaging; its decay product, technetium-99m ( 99 mTc), is a metastable isotope with a half-life of ~6 hours. Such a short half-life does not allow for production, separation, and application to a patient before decaying further. Therefore, the solution is to produce the precursor to 99 mTc, 99 Mo. 99 Mo has a half-life of 66 hours, long enough to produce the material and send it to the desired location before it fully decays. Under National Nuclear Security Administration’s (NNSA) highly enriched uranium (HEU) minimization mission, the Material Management and Minimization (M3) office leads the Molybdenum-99 (Mo-99) program. A major goal of the Mo-99 program is to develop methods of producing 99 Mo using low enriched uranium (LEU) opposed to the traditionally used HEU domestically in the U.S. Internationally, LEU is the new standard for use in research reactors and isotope production facility. In this study, we introduce a generic fissile solution system for 99 Mo production facility meeting the design requirement of 1) LEU usage as fissile material and 2) subcritical fissile system coupled with a horizontal electron beam (E-beam) accelerator. The focus of this research is to optimize a fissile solution configuration for maximum 99 Mo production yield using MCNP model, and ultimately implement the optimized design with computational fluid dynamics (CFD) model to understand thermal hydraulic behavior and solution convection characteristic. The report consists of two parts: 1) isotope yield calculation and criticality analysis with a wide range of parametric test matrix, and 2) thermal hydraulic analysis for solution mixing and cooling assessment.

60 APPLIED LIFE SCIENCES↗
NASA NTRS1985

Elemental technetium as a cosmic-ray clock

Several radioactive isotopes have been proposed as clocks for the study of the mean cosmic ray confinement time, T sub e. Measurements of Be-10 and Al-26 give a value for T sub e of about 10 Myr when one uses a leaky box cosmic ray propagation model. It is important to obtain additional measurements of T sub e from other radioactive isotopes in order to check whether the confinement is the same throughout the periodic table. The possible use of Tc (Z = 43) as a cosmic clock is investigated. Since all isotopes of Tc are radioactive, one might be able to group these isotopes and use the elemental abundance as a whole. The results of the calculations are somewhat inconclusive for two reasons. First, the beta + decay half lives of two of the Tc isotopes relevant to our calculation are not known. Second, the dependence of the Tc abundance on the mean confinement time is rather weak when one considers the number of events expected in 4 trays of plastic track detectors. However, a future, finite measurement of the Beta + half lives and the possible use of the entire collecting area of the HNC to detect Tc nuclei could make the use of Tc as a cosmic ray clock more attractive.

Drach, J.↗
DOE OSTI2016

Design of a Commercial Scale Accelerator Driven Subcritical Aqueous Assembly

SHINE Medical Technologies is working towards producing a reliable, domestic supply of medical isotopes, including molybdenum-99 (Mo-99), xenon-133, and iodine 131, without using highly enriched uranium. Approximately 88% of Mo-99 decays to technetium-99m, which is the most widely used medical isotope in the world, accounting for approximately 80% of all nuclear medicine procedures. A novel aqueous, subcritical nuclear assembly has been designed by SHINE Medical Technologies. The assembly can produce medical radioisotopes in significant quantities for U.S. and foreign needs. The assembly produces isotopes without using a reactor, instead using a fusion-based neutron source coupled to a fission-based aqueous assembly. The subcritical nature of the aqueous assembly has safety and licensing benefits, including the elimination of the reactor control system and elimination of high level waste. SHINE plans to construct eight subcritical assemblies with a plant capacity of greater than 3000 6-day Curies, which is more than one-quarter of the world's Mo-99 requirements. From a nuclear design standpoint, the aqueous subcritical assembly needs to be able to generate sufficient fission power to meet the production capacity requirements while remaining subcritical at all times with an acceptable margin. To address these conflicting design goals, a simple nuclear design and reactivity control philosophy has been developed that allows for high power operation with a high degree of safety. A fusion-driven subcritical fission assembly has been designed by SHINE Medical Technologies to provide a reliable, domestic supply of medical isotopes. Nuclear models have been created and show that the assembly can be started up, operated, and shutdown in a safe manner without active reactivity control systems. A unique reactivity control strategy allows SHINE to operate the aqueous subcritical assembly at much higher power levels than would otherwise be practical, while still retaining the benefits of subcritical operation and avoiding the need for complex subcritical measurement techniques. Based on the current design, the subcritical target is capable of generating significant fission power (on the order of 100 kW) when being driven by a simple, low-cost deuterium-tritium accelerator, enabling the SHINE facility to supply over one-fourth of the worldwide demand of Mo- 99. SHINE has recently been issued a construction permit by the U.S. Nuclear Regulatory Commission to build the commercial facility in Janesville, WI, with eight subcritical assemblies and the associated production equipment. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Distribution of Prostate Sentinel Nodes: A SPECT-Derived Anatomic Atlas

Purpose: The randomized Radiation Therapy Oncology Group 94-13 trial revealed that coverage of the pelvic lymph nodes in high-risk prostate cancer confers an advantage (progression-free survival and biochemical failure) in patients with {>=}15% risk of lymph node involvement. To facilitate an improved definition of the adjuvant target volume, precise knowledge regarding the location of the relevant lymph nodes is necessary. Therefore, we generated a three-dimensional sentinel lymph node atlas. Methods and Materials: In 61 patients with high-risk prostate cancer, a three-dimensional visualization of sentinel lymph nodes was performed using a single photon emission computed tomography system after transrectal intraprostatic injection of 150 to 362 (median 295) mega becquerel (MBq) {sup 99m}Technetium-nanocolloid (1.5-3h after injection) followed by an anatomic functional image fusion. Results: In all, 324 sentinel nodes in 59 of 61 patients (96.7%) were detected, with 0 to 13 nodes per patient (median 5, mean 5.3). The anatomic distribution of the sentinel nodes was as follows: external iliac 34.3%, internal iliac 17.9%, common iliac 12.7%, sacral 8.6%, perirectal 6.2%, left paraaortic 5.3%, right paraaortic 5.3%, seminal vesicle lymphatic plexus 3.1%, deep inguinal 1.5%, superior rectal 1.2%, internal pudendal 1.2%, perivesical 0.9%, inferior rectal 0.9%, retroaortic 0.3%, superficial inguinal 0.3%, and periprostatic 0.3%. Conclusions: The distribution of sentinel nodes as detected by single photon emission computed tomography imaging correlates well with the distribution determined by intraoperative gamma probe detection. A lower detection rate of sentinels in close proximity to the bladder and seminal vesicles is probably caused by the radionuclide accumulation in the bladder. In regard to intensity-modulated radiotherapy techniques, the presented anatomic atlas may allow optimized target volume definitions.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2011

Technical Solutions to Ensure Safe Yttrium-90 Radioembolization in Patients With Initial Extrahepatic Deposition of {sup 99m}Technetium-Albumin Macroaggregates

Purpose: To evaluate the incidence of extrahepatic deposition of technetium-99m-labeled albumin macroaggregates ({sup 99m}Tc-MAA) after pretreatment angiography, before yttrium-90 radioembolizaton ({sup 90}Y-RE), and to report on technical solutions that can be used to ensure safe delivery of {sup 90}Y-microspheres in patients with initial extrahepatic deposition. Materials and Methods: A retrospective analysis of 26 patients with primary and secondary liver malignancies, who were scheduled for treatment with {sup 90}Y-RE in our institution in 2009, was performed. The angiograms and single-photon emission computed tomography images of all patients were reviewed by an interventional radiologist and a nuclear medicine physician, respectively, to identify and localize extrahepatic deposition of {sup 99m}Tc-MAA when present. Subsequently, the technical solutions were used to successfully perform {sup 90}Y-RE in these patients were evaluated and described. Results: Extrahepatic deposition of {sup 99m}Tc-MAA was observed in 8 of 26 patients (31%). In 7 of 8 patients, a second pretreatment angiography was performed to detect the cause of extrahepatic deposition. The technical solutions to enable safe {sup 90}Y microspheres delivery included more distal placement of the microcatheter in the proper/right hepatic artery in 4 of 7 (57%) patients; (super)selective catheterization of multiple segmental branches in 2 of 7 (29%); and additional coiling of a newly detected branch in the remaining patient (14%). This was confirmed by a second MAA procedure. {sup 90}Y-RE was eventually performed in 25 of 26 (96%) patients. No procedure-related complications (<30 days) were observed. Conclusion: Extrahepatic deposition of {sup 99m}Tc-MAA after pretreatment angiography did occur in 8 of 26 (31%) patients. The technical solutions as presented allowed safe {sup 90}Y-RE delivery in 25 of 26 (96%) patients.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2025

Evaluate opportunities for monitoring the oxidation states of technetium in both organic and aqueous phases using Raman spectroscopy

An important isotope in the nuclear fuel cycle, 99 Tc is characterized by its long half-life of 2.1 × 10 5 years and its beta emission. It is generated through the fission of 235 U. As the burnup levels of nuclear waste continue to rise, the concentration of technetium in spent fuel increases, leading to a heightened interest in its quantification during the reprocessing of spent nuclear fuel. In HNO3 environments, technetium predominantly exists as TcO 4 − ; however, it tends to interfere with reducing agents, decreasing the separation efficiency for actinides. Therefore, accurate quantitative determination of technetium is critical to controlling its distribution and mitigating its negative effect on the reprocessing of spent fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Preliminary Analysis of Species Partitioning in the DWPF Melter - Sludge Batch 7b

The goal of the work described in this report is to provide additional off-gas carryover data from the Defense Waste Processing Facility (DWPF) melter to support the estimation of off-gas carryover rates from the Low Activity Waste (LAW) melter at the Hanford Site. Of particular interest is to determine the volatility of technetium-99 (Tc-99), which is a key isotope of concern for disposal and is a significant challenge to retain in glass in the Hanford waste treatment flowsheet. Since the DWPF melter is not equipped with routine offgas monitoring and/or sampling capabilities, the off-gas carryover data had to be extracted using component mass balance of the measured feed and pour stream (PS) glass compositions under time-averaged melter operating conditions over the duration of one canister-filling cycle, typically 20-25 hours. The method was first tested using the Sludge Batch 6 (SB6) PS sample data taken while Canister #3472 was being filled on 12/20/2010, approximately three months after the (argon) glass bubblers were installed. The method was tested again using the SB7a PS sample data taken while Canister #3619 was being filled on 8/5/2011 with the glass bubblers in operation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Technetium-99M generator for enriched molybdenum

A method for separating a parent isotope from a daughter isotope is provided, the method comprising supplying irradiated target; dissolving the irradiated target; treating the dissolved irradiated target to a precipitation step to form a first solid phase of the parent isotope and a first liquid phase of the daughter isotope; filtering the first liquid phase of daughter isotope to create a first purified fraction of the daughter isotope to create a second solid phase of the parent isotope and a second liquid phase of the daughter isotope; and filtering the second liquid phase to create a second liquid fraction of the daughter isotope. The process can be repeated until the 99 Mo is exhausted and the enriched target material 98 Mo or 100 Mo can be easily re-used.

Brown, Michael A.↗
DOE OSTI2018

The ARGUS Solution Reactor and Molybdenum Production: A Summary Report Based on Open Literature

The (Aqueous Homogeneous Reactor) AHR was conceptualized at Los Alamos for work in advancing weapon design with initial experiments completed at Los Alamos National Laboratory (LANL) and Oak Ridge National Laboratory (ORNL) in the early 1940s and 1950s. The reactor design was discovered to be an efficient mean of production for medical radio isotopes (e.g. molybdenum-99 and other fission products) due to its streamlined process of extracting target radioactive sources from the fissile solution via a separation system. As the demand for medical radioisotopes has increased so has the need for a safe and low cost isotope source, the most commonly used of which is technetium-99m (99m Tc), daughter product from molybdenum decay. Studies indicate solution reactors pose a significantly lower hazard to the workers, public and environment, compared to other modern reactors (International Atomic Energy Agency 2008). The inherent safety and ability to selectively extract radionuclides from its fuel solution make AHRs particularly well suited for medical isotope production. This discussion narrows its focus to one of the longest running aqueous homogeneous reactor systems, ARGUS. A study of the conversion of ARGUS to a low-enriched uranyl (LEU) sulfate fuel solution was part of the Global Threat Reduction Initiative (GTRI) program. The goal of this paper is to provide a summary of the ARGUS reactor system based on open literature available.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2015

A biokinetic model for systemic technetium in adult humans

The International Commission on Radiological Protection (ICRP) currently is updating its biokinetic and dosimetric models for internally deposited radionuclides. Technetium (Tc), the lightest element that exists only in radioactive form, has two important isotopes from the standpoint of potential risk to humans: the long-lived isotope 99 Tm(T 1/2 =2.1x10 5 y) is present in high concentration in nuclear waste, and the short-lived isotope 99m Tc (T 1/2 =6.02 h) is the most commonly used radionuclide in diagnostic nuclear medicine. This paper reviews data on the biological behavior of technetium and proposes a biokinetic model for systemic technetium in the adult human body for use in radiation protection. Compared with the ICRP s current occupational model for systemic technetium, the proposed model provides a more realistic description of the paths of movement of technetium in the body; provides greater consistency with experimental and medical data; and, for most radiosensitive organs, yields substantially different estimates of cumulative activity (total radioactive decays within the organ) following uptake of 99 Tm or 99 mTc to blood.

62 RADIOLOGY AND NUCLEAR MEDICINE↗