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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Direct Down-scale Experiments of Concentration Column Designs for SHINE Process

Argonne is assisting SHINE Medical Technologies in their efforts to become a domestic Mo-99 producer. The SHINE accelerator-driven process uses a uranyl-sulfate target solution for the production of fission-product Mo-99. Argonne has developed a molybdenum recovery and purification process for this target solution. The process includes an initial Mo recovery column followed by a concentration column to reduce the product volume from 15-25 L to < 1 L prior to entry into the LEU Modified Cintichem (LMC) process for purification.1 This report discusses direct down-scale experiments of the plant-scale concentration column design, where the effects of loading velocity and temperature were investigated.

42 ENGINEERING↗
DOE OSTI2016

Production of Low Enriched Uranium Nitride Kernels for TRISO Particle Irradiation Testing

A large batch of UN microspheres to be used as kernels for TRISO particle fuel was produced using carbothermic reduction and nitriding of a sol-gel feedstock bearing tailored amounts of low-enriched uranium (LEU) oxide and carbon. The process parameters, established in a previous study, produced phasepure NaCl structure UN with dissolved C on the N sublattice. The composition, calculated by refinement of the lattice parameter from X-ray diffraction, was determined to be UC 0.27 N 0.73 . The final accepted product weighed 197.4 g. The microspheres had an average diameter of 797±1.35 μm and a composite mean theoretical density of 89.9±0.5% for a solid solution of UC and UN with the same atomic ratio; both values are reported with their corresponding calculated standard error.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Scoping study to expedite development of a field deployable and portable instrument for UF6 enrichment assay

The primary objective of the present study is to identity the most promising, viable technologies that are likely to culminate in an expedited development of the next-generation, field-deployable instrument for providing rapid, accurate, and precise enrichment assay of uranium hexafluoride (UF6). UF6 is typically involved, and is arguably the most important uranium compound, in uranium enrichment processes. As the first line of defense against proliferation, accurate analytical techniques to determine the uranium isotopic distribution in UF6 are critical for materials verification, accounting, and safeguards at enrichment plants. As nuclear fuel cycle technology becomes more prevalent around the world, international nuclear safeguards and interest in UF6 enrichment assay has been growing. At present, laboratory-based mass spectrometry (MS), which offers the highest attainable analytical accuracy and precision, is the technique of choice for the analysis of stable and long-lived isotopes. Currently, the International Atomic Energy Agency (IAEA) monitors the production of enriched UF6 at declared facilities by collecting a small amount (between 1 to 10 g) of gaseous UF6 into a sample bottle, which is then shipped under chain of custody to a central laboratory (IAEA’s Nuclear Materials Analysis Laboratory) for high-precision isotopic assay by MS. The logistics are cumbersome and new shipping regulations are making it more difficult to transport UF6. Furthermore, the analysis is costly, and results are not available for some time after sample collection. Hence, the IAEA is challenged to develop effective safeguards approaches at enrichment plants. In-field isotopic analysis of UF6 has the potential to substantially reduce the time, logistics and expense of sample handling. However, current laboratory-based MS techniques require too much infrastructure and operator expertise for field deployment and operation. As outlined in the IAEA Department of Safeguards Long-Term R&D Plan, 2012–2023, one of the IAEA long-term R&D needs is to “develop tools and techniques to enable timely, potentially real-time, detection of HEU (Highly Enriched Uranium) production in LEU (Lowly Enriched Uranium) enrichment facilities” (Milestone 5.2). Because it is common that the next generation of analytical instruments is driven by technologies that are either currently available or just now emerging, one reasonable and practical approach to project the next generation of chemical instrumentation is to track the recent trends and to extrapolate them. This study adopted a similar approach, and an extensive literature review on existing and emerging technologies for UF6 enrichment assay was performed. The competitive advantages and current limitations of different analytical techniques for in-field UF6 enrichment assay were then compared, and the main gaps between needs and capabilities for their field use were examined. Subsequently, based on these results, technologies for the next-generation field-deployable instrument for UF6 enrichment assay were recommended. The study was organized in a way that a suite of assessment metric was first identified. Criteria used in this evaluation are presented in Section 1 of this report, and the most important ones are described briefly in the next few paragraphs. Because one driving force for in-field UF6 enrichment assay is related to the demanding transportation regulation for gaseous UF6, Section 2 contains a review of solid sorbents that convert and immobilized gaseous UF6 to a solid state, which is regarded as more transportation friendly and is less regulated. Furthermore, candidate solid sorbents, which show promise in mating with existing and emerging assay technologies, also factor into technology recommendations. Extensive literature reviews on existing and emerging technologies for UF6 enrichment assay, covering their scientific principles, instrument options, and current limitations are detailed in Sections 3 and 4, respectively. In Section 5, the technological gaps as well as start-of-the-art and commercial off-the-shelf components that can be adopted to expedite the development of a fieldable or portable UF6 enrichment-assay instrument are identified and discussed. Finally, based on the results of the review, requirements and recommendations for developing the next-generation field-deployable instrument for UF6 enrichment assay are presented in Section 6.

36 MATERIALS SCIENCE↗
DOE OSTI2017

HFIR Fuel Casting Support

Process exploration for fuel production for the High Flux Isotope Reactor (HFIR) using cast LEU-10wt.%Mo as an initial processing step has just begun. This project represents the first trials concerned with casting design and quality. The studies carried out over the course of this year and information contained in this report address the initial mold development to be used as a starting point for future operations. In broad terms, the final billet design is that of a solid rolling blank with an irregular octagonal cross section. The work covered here is a comprehensive view of the initial attempts to produce a sound casting. This report covers the efforts to simulate, predict, cast, inspect, and revise the initial mold design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Materials and Methods for Streamlined Laboratory Analysis of Environmental Samples, FY 2016 Report

The International Atomic Energy Agency (IAEA) relies upon laboratory analysis of environmental samples (typically referred to as “swipes”) collected during on-site inspections of safeguarded facilities to support the detection and deterrence of undeclared activities. Unfortunately, chemical processing and assay of the samples is slow and expensive. A rapid, effective, and simple extraction process and analysis method is needed to provide certified results with improved timeliness at reduced costs (principally in the form of reduced labor), while maintaining or improving sensitivity and efficacy. To address these safeguard needs the Pacific Northwest National Laboratory (PNNL) explored and demonstrated improved methods for environmental sample (ES) analysis. Improvements for both bulk and particle analysis were explored. To facilitate continuity and adoption, the new sampling materials and processing methods will be compatible with existing IAEA protocols for ES analysis. PNNL collaborated with Oak Ridge National Laboratory (ORNL), which performed independent validation of the new bulk analysis methods and compared performance to traditional IAEA’s Network of Analytical Laboratories (NWAL) protocol. ORNL efforts are reported separately. This report describes PNNL’s FY 2016 progress, which was focused on analytical application supporting environmental monitoring of uranium enrichment plants and nuclear fuel processing. In the future the technology could be applied to other safeguard applications and analytes related to fuel manufacturing, reprocessing, etc. PNNL’s FY 2016 efforts were broken into two tasks and a summary of progress, accomplishments and highlights are provided below. Principal progress and accomplishments on Task 1, Optimize Materials and Methods for ICP-MS Environmental Sample Analysis, are listed below. • Completed initial procedure for rapid uranium extraction from ES swipes based upon carbonate-peroxide chemistry (delivered to ORNL for evaluation). • Explored improvements to carbonate-peroxide rapid uranium extraction chemistry. • Evaluated new sampling materials and methods (in collaboration with ORNL). • Demonstrated successful ES extractions from standard and novel swipes for a wide range uranium compounds of interest including UO 2 F 2 and UO 2 (NO 3 ) 2 , U 3 O 8 and uranium ore concentrate. • Completed initial discussions with commercial suppliers of PTFE swipe materials. • Submitted one manuscript for publication. Two additional drafts are being prepared. Principal progress and accomplishments on Task 2, Optimize Materials and Methods for Direct SIMS Environmental Sample Analysis, are listed below. • Designed a SIMS swipe sample holder that retrofits into existing equipment and provides simple, effective, and rapid mounting of ES samples for direct assay while enabling automation and laboratory integration. • Identified preferred conductive sampling materials with better performance characteristics. • Ran samples on the new PNNL NWAL equivalent Cameca 1280 SIMS system. • Obtained excellent agreement between isotopic ratios for certified materials and direct SIMS assay of very low levels of LEU and HEU UO 2 F 2 particles on carbon fiber sampling material. Sample activities range from 1 to 500 CPM (uranium mass on sample is dependent upon specific isotope ratio but is frequently in the subnanogram range). • Found that the presence of the UF molecular ions, as measured by SIMS, provides chemical information about the particle that is separate from the uranium isotopics and strongly suggests that those particles originated from an UF6 enrichment activity. • Submitted one manuscript for publication. Another manuscript is in preparation.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2017

Verification of Historical System Transient Simulations for the RHF Research Reactor

The RELAP5/Mod 3.3 [1] reactor transient analysis code was used to perform a verification of the RHF reactor system transients previously simulated with the CATHARE Code. The verification included comparisons of temperature, pressure, flows, etc. where direct code comparisons could be made. This work is a precursor to the simulations that will be performed with the RELAP5 code to evaluate the impact of converting the RHF reactor from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Analysis of Radiolytically Generated Gases in Mini-AMORE Experiment

Argonne National Laboratory (Argonne) is assisting one of the potential domestic Mo-99 producers, SHINE Medical Technologies, as part of the National Nuclear Security Administration’s (NNSA) Office of Material, Management, and Minimization (M 3 ) program in the development of a domestic Mo-99 production pathway that does not use highly enriched uranium (HEU). The process proposed by SHINE would produce Mo-99 by neutron-induced fission of a low enriched uranium (LEU) solution as uranyl sulfate in a subcritical assembly. As part of the development effort, Argonne is undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Mini-AMORE is a part of the AMORE experiment where 2 mL samples of uranyl sulfate solution are inserted in the dry well of the AMORE target solution vessel.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2018

Historical Review of Stress Corrosion Cracking in Concentrated Uranium-Molybdenum Alloys

This report presents a historical overview of stress corrosion cracking (SCC) in uranium-molybdenum alloys, focusing primarily on U-10 wt. pct. Mo (U-10Mo). This alloy has been chosen as the high density uranium monolithic fuel for converting research reactors from high enriched uranium (HEU) to low enriched uranium (LEU) fuel for its combination of fuel density and gamma phase stability. One disadvantage of U-10Mo is its propensity towards stress corrosion cracking (SCC), driven by high residual stresses resulting from its high yield strength. This report contains a metallurgical overview of U-10Mo alloys incorporating processing, microstructural evolution, and mechanical properties. This report also summarizes SCC literature including contributing factors, initiation and propagation, the mechanism responsible for U-10Mo SCC, and possible solutions to the problem. A brief synopsis of interviews with key personnel from several Department of Energy (DOE) facilities and BWX Technologies (BWXT) is also included. Recommendations for future studies are outlined.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

AGR 3/4 Irradiation Test Final As Run Report, Revision 1

Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5 × 10 25 n/m 2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the temperature in the twelve capsules relatively flat in a range of temperatures suitable for the measurement of fission product diffusion in compact matrix and structural graphite materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Prototype Stilbene Neutron Collar II

A neutron collar using stilbene organic scintillator cells for fast neutron counting is described for the assay of fresh low enriched uranium (LEU) fuel assemblies. The prototype stilbene collar has a form factor similar to standard He-3 based collars and uses an AmLi interrogation neutron source. This report describes the simulation, using a custom LLNL developed MCNPX, of list mode neutron correlation data on various fuel assemblies. Some of these fuel assemblies have neutron absorbers (burnable Gd poisons) and partial defect s (DU, Fe, Pb pins). Calibration curves (doubles vs 235 U linear mass density) are presented for fast (with Cd lining) mode of operation. The modeling incorporates LLNL developed neutron crosstalk correction algorithms and stilbene light output detector response quench function. Pileup issues are addressed using the LLNL code GamGen / RadSrc to estimate the flux of alpha decay gammas from the fuel assemblies on the stilbene cells with and without Pb shielding. One panel of the prototype collar using 10 high-efficiency 4”x2” stilbene cells from Inrad Optics, using LLNL developed technology, has been constructed and calibrated for optimal PSD and data acquisition electronics. Several sources (AmLi, AmBe, Cf, DD, and DT) have been measured with the single panel and the measured data is consistent with the modeling. A novel part of our work was to determine the light output (LO) quench function for the Inrad 4” x 2” stilbene crystals through a LLNL-MCNPX simulation based optimized fitting of the measured LO data.

42 ENGINEERING↗
DOE OSTI2018

Unreviewed Disposal Question Evaluation: Disposal of the Nevada National Security Site Uranium Powder at the Area 5 Radioactive Waste Management Site, Nevada National Security Site

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Nevada National Security Site (NNSS) Uranium Powder (Normal, depleted uranium [DU], low-enriched uranium [LEU]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 OSTI2018

Historical Review of Stress Corrosion Cracking in Concentrated Uranium-Molybdenum Alloys

This report presents a historical overview of stress corrosion cracking (SCC) in uranium-molybdenum alloys, focusing primarily on bare U-10 wt. pct. Mo (U-10Mo). The fuel system being pursued by the United States High Performance Research Reactor (USHPRR) project incorporates a zirconium interlayer applied early in the process which likely serves as a protective coating. This alloy has been chosen as the high density uranium monolithic fuel for converting research reactors from high enriched uranium (HEU) to low enriched uranium (LEU) fuel for its combination of fuel density and gamma phase stability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Channel Gap Probe EMPIrE Report

EMPIrE aims to provide data for the "comprehension phase" of the European Low Enriched Uranium (LEU ) U-Mo Fuel Development program for the conversion of European civilian research reactors. The majority of the fuel performance data will be obtained during post-irradiation examinations (PIE) of the fuel. The primary EMPIrE fuel specimen design is based on uranium-molybdenum (U-Mo) alloy particles dispersed in an aluminum matrix clad in the European alloy AG3NE and in the form of mini-plate sized specimens (1" x 4" plates). Secondary EMPIrE specimens include U Mo monolithic fuel clad in European alloy AlFeNi. The experiment is designed in such a way that U-Mo samples achieve three distinct end-of-life burnup groupings. Irradiation has taken place at ATR (Advance Treatment Reactor). The fuel plates are contained in fuel capsules holding four plates each, with 2 plates at 2 levels. Once the capsules with the plates are irradiated, they have to be inspected during the outage in-between cycles. The capsules need to have visual inspection performed and channel gap widths measured prior to reinsertion. These inspections are to ensure that the plates have not deformed significantly. Obstruction of the coolant channels and restriction of the coolant flow are less of a concern in EMPIrE testing as the channels were made intentionally wide. The Channel Gap Probe (CGP) has been developed to measure channel gap widths, which serves as indirect measures of the fuel plate performance and deformation in the course of irradiation which enables the evaluation of the risk from continuing irradiation prior to reinsertion of the plates in reactor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Experimental Evaluation of the Fast-Neutron Passive Collar (FNPC) Performance (Interim Report)

This report represents a deliverable in the Safeguards Technology WBS # Project Title, 24.1.3.1, Task 3 (Report on laboratory-based performance testing of the FNPC). A new, simplified method for the verification of the 235 U mass in fresh LEU fuel assemblies that is relatively independent of burnable poison content is being developed. LWR fresh fuel is currently verified using the uranium neutron coincidence collar (UNCL), which relies on an AmLi source to induce fission in the 235 U. AmLi neutrons are slowed down in polyethylene, so the induced fission is primarily due to thermal neutrons. Burnable poisons, which are added to nuclear fuel assemblies to extend the lifetime of the fuel, absorb neutrons that would otherwise have induced fission in the fuel, thereby reducing the neutron count rate and thus the measured 235 U mass. To remove this bias factor a correction is applied based on the operator’s declared burnable poison content. The new technique uses 238 U spontaneous fission in LWR fresh fuel rods to self-interrogate the 235 U mass. The 238 U spontaneous fission neutrons have a hard neutron energy spectrum, which means that the technique is less sensitive to burnable poisons than the AmLi based systems. This project has continued the work done over the past two years to study this new technique using the optimized 3 He based fastneutron passive collar (FNPC). Fabrication of the FNPC has been completed and measurements are in progress using the LANL mockup PWR fuel assembly. Also, preliminary evaluations of advanced analysis techniques and unattended mode operation have been performed. In FY18, the 3 He tube based detector was compared with a 10 B sealed cell based detector for the same set of PWR fuel assemblies in the Rodeo-II program. The results were that the optimized 3 He detector had higher efficiency and less weight than the 10 B based system where both systems had about the same cost. Thus, the FY19 work continued exclusively with the 3 He based detector; although, either system had the potential for future applications. However, the high efficiency of the FPNC (~24%) is critical for precise measurements of the triples rate necessary in the multiplicity advanced analysis technique which can solve for the burnable poison content.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Integral Experiment Request 209 (CED-2 Summary Report)

This report examines proposed Seven Percent Critical Experiment (7uPCX) experiments with fuel arrays larger than would be critical when fully reflected. In these experiments, the reactivity of the assembly will be controlled by varying the moderator/reflector level in the core tank. The analysis uses two configurations, each completely filling the 45x45 fuel rod array with fuel rods and water holes, as representative examples of the proposed experiments. The proposed configurations are compared to the experiments documented in LEU-COMP-THERM-078 [Reference 1] and to fully-reflected experiments with the same fully-loaded fuel arrays that are poisoned with boron in the moderator. The conclusion is drawn that the proposed experiments can be performed with acceptably low uncertainties given a calibrated moderator/reflector level measurement system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Integral Experiment Request 206 CED-3a Summary Report

Under IER-206, the hardware for the Burnup Credit Critical Experiment will be removed from storage and placed in operation. This is the hardware we used in 2002 to perform the experiments that eventually documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments as LEU-COMP-THERM-079.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗