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At least 1,297 records · Page 72
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Metal Matrix Microencapsulated (M3) fuel neutronics performance in PWRs

Metal Matrix Microencapsulated (M3) fuel consists of TRISO or BISO coated fuel particles directly dispersed in a matrix of zirconium metal to form a solid rod (Fig. 1). In this integral fuel concept the cladding tube and the failure mechanisms associated with it have been eliminated. In this manner pellet-clad-interactions (PCI), thin tube failure due to oxidation and hydriding, and tube pressurization and burst will be absent. M3 fuel, given the high stiffness of the integral rod design, could as well improve grid-to-rod wear behavior. Overall M3 fuel, compared to existing fuel designs, is expected to provide greatly improved operational performance. Multiple barriers to fission product release (ceramic coating layers in the coated fuel particle and te metal matrix) and the high thermal conductivity zirconium alloy metal matrix contribute to the enhancement in fuel behavior. The discontinuous nature of fissile material encapsulated in coated particles provides additional assistance; for instance if the M3 fuel rod is snapped into multiple pieces, only the limited number of fuel particles at the failure cross section are susceptible to release fission products. This is in contrast to the conventional oxide fuel where the presence of a small opening in the cladding provides the pathway for release of the entire inventory of fission products from the fuel rod. While conventional metal fuels (e.g. U-Zr and U-Mo) are typically expected to experience large swelling under irradiation due to the high degree of damage from fission fragments and introduction of fission gas into the lattice, this is not the case for M3 fuels. The fissile portion of the fuel is contained within the coated particle where enough room is available to accommodate fission gases and kernel swelling. The zirconium metal matrix will not be exposed to fission products and its swelling is known to be very limited when exposed solely to neutrons. Under design basis RIA and LOCA, fuel performance will be superior to the conventional oxide fuel since PCMI and rod burst, respectively, have been mitigated. Under beyond design basis accident scenarios where the fuel is exposed to high temperature steam for prolonged periods, larger inventory of zirconium metal in the core could negatively affect the accident progression. A thin steam resistant layer (e.g. alumina forming alloy steel), integrated into the solid rod during fabrication by co-extrusion or hot-isostatic-pressing, offers the potential to provide additional fuel protection from steam interaction: blanketing under a range of boiling regimes and under severe accident conditions up to high temperatures well beyond what is currently possible in the conventional fuel. A crucial aspect to the viability of M3 fuel in light water reactors is the reduced heavy metal load compared to standard pellet fuel. This study evaluated the design requirements to operate a Pressurized Water Reactor (PWR) with M3 fuel in order to obtain fuel cycle length, reactivity coefficients, and power peaking factors comparable to that of standard fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

EOS & damage behavior of S200-F Be

We have performed a series of plate impact experiments on both vacuum hot-pressed and Roll-textured S-200F Be on the 40 mm gun at TA-55. This presentation provides an update of the progress to date for this effort.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Verifying disarmament: scientific, technological and political challenges

There is growing interest in, and hopes for, nuclear disarmament in governments and nongovernmental organizations (NGOs) around the world. If a nuclear-weapon-free world is to be achievable, verification and compliance will be critical. VerifYing disarmament would have unprecedented scientific, technological and political challenges. Verification would have to address warheads, components, materials, testing, facilities, delivery capabilities, virtual capabilities from existing or shutdown nuclear weapon and existing nuclear energy programs and material and weapon production and related capabilities. Moreover, it would likely have far more stringent requirements. The verification of dismantlement or elimination of nuclear warheads and components is widely recognized as the most pressing problem. There has been considerable research and development done in the United States and elsewhere on warhead and dismantlement transparency and verification since the early 1990s. However, we do not today know how to verifY low numbers or zero. We need to develop the needed verification tools and systems approaches that would allow us to meet this complex set of challenges. There is a real opportunity to explore verification options and, given any realistic time frame for disarmament, there is considerable scope to invest resources at the national and international levels to undertake research, development and demonstrations in an effort to address the anticipated and perhaps unanticipated verification challenges of disarmament now andfor the next decades. Cooperative approaches have the greatest possibility for success.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2020

Elevated-temperature cyclic properties of advanced manufactured materials

Microreactor developers at the 2019 GAIN Microreactors workshop expressed a need for advanced manufacturing to fabricate microreactor components. Powder metallurgy (PM) hot isostatic pressing (HIP) is more mature than other advanced manufacturing techniques for the following reasons: 1. a code case exists for Section III, Division 1, Subsection NB components of light water reactors to be manufactured using PM HIP Alloy 316L, and 2. tensile and creep properties have been shown to be equivalent to wrought material for Type 316 stainless steel and Grade 91. Thus, PM HIP is more readily deployable in the near term. Elevated-temperature cyclic properties of PM-HIP materials, however, have not been established. The scope of testing for Code qualification could be significantly reduced compared to a new material for the following reasons: 1. only a limited amount of time consuming creep testing would be required, and 2. testing of fatigue and creep-fatigue behavior could be reduced if it can be demonstrated that the properties are comparable to wrought material. The objective of this work package is to develop an understanding of the relationship of advanced materials processing on the material characteristics as they relate to elevated-temperature component design and construction. This will be achieved by the following: • Elevated-temperature fatigue and creep-fatigue testing of PM HIP and wrought Alloy 316L. • Fracture and microstructural characterization of the PM HIP and wrought Alloy 316L. • Analysis of the elevated-temperature fatigue and creep-fatigue properties of PM HIP Alloy 316L. • Draft a report evaluating the elevated-temperature fatigue and creep-fatigue properties of PM HIP Alloy 316L. • Pursue procurement of Alloy 316H and Grade 91 manufactured using PM HIP. • Initiate scoping studies of the procured Alloy 316H or Grade 91 manufactured using PM HIP contingent upon successful material procurement.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Parameter survey on heating conditions for glass-bonded sodalite ceramic waste fabrication from type-A zeolite containing simulating FPs

In the pyrometallurgical reprocessing for metal fuel cycle, fission products (FPs) in the spent salt is converted into a glass-bonded sodalite by means of Pressure-less Consolidation (PC). Although a standard condition of the PC method was reported, the reason for choosing this condition is not clear particularly for the temperature condition. In the present study, a parameter survey on the heating condition was performed by fabricating the glass-bonded sodalite containing simulating FPs under various conditions. The maximum temperature, the heating duration, the glass ratio in the initial material, and the weight load for pressing the material were chosen as the variable parameters. The mass balance of each element and the quality of the product were evaluated. It was exhibited that both the volatilized content and the free salt in the product were reduced by lowering the maximum temperature. The apparent density of the product was enhanced by both increasing the heating duration and increasing the weight load. Accordingly, lowering the maximum temperature while increasing the weight load was chosen to modify the fabricating condition. As a result of the modified condition, both the volatilized and the free salt ratios were reduced without significant change in the apparent density of the product. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fabrication of rodlets with coated and lined Mo-alloy cladding for testing and irradiation

Molybdenum alloy fuel cladding with an outer protective layer has been proposed to achieve enhanced tolerance to severe loss of coolant accidents for light water reactors. The accident tolerant fuel cladding will need to meet stringent fuel reliability requirements and significantly delay steam reaction during a severe accident to provide plant operators additional coping time to recover from the accident. A laboratory measurable target for this feasibility study with coated Mo cladding is to demonstrate its capability to survive in steam and maintain fuel rods in coolable geometry at 1200-1500 Celsius degrees for at least 24 hours. Preliminary studies indicate such a capability may provide additional coying of 4-19 hours. A feasibility study undertaken since 2012 has demonstrated feasibility of fabricating long, thin-wall Mo and Mo-ODS tubes, and forming a metallurgically bonded protective coating of FeCrAl or Zr-alloy via physical vapor deposition to provide protection of the Mo cladding in 1200 C. degrees for 24 hours. The diametral ductility of the Mo and Mo-ODS thin wall tubes can be increased to 5-15% pre-irradiation to meet fuel operational requirements. Fabrication of the outer protective layer using thermo-mechanical processing, such as hot isostatic pressing and hot coextrusion, has been under evaluation to achieve high quality lined Mo cladding from commercial scale production. Welding of lined Mo tubes to end-caps are being developed using a combination of solid state welding of laser or electron beam seal welding. Rodlets of lined Mo tubes are being fabricated for further testing to validate its reliability under normal operation and capability of accident tolerance to the 1200- 1500 C. degrees range. Test reactor irradiation is also being planned. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Study of the U{sub 3}O{sub 8} effect on MOX fuel sintering mechanisms

The manufacturing of uranium-plutonium mixed oxides fuel (also known as MOX) for light water reactor (LWR), starting generally from UO{sub 2} and PuO{sub 2} raw powders and resulting in (U,Pu)O{sub 2} pellets, involves several steps: i) preparation of a UO{sub 2}+PuO{sub 2} master blend with a Pu content of about 25-30 at% by milling and sieving, ii) dilution by mixing the master blend with non-milled UO{sub 2} to reach a Pu content of about 10 at%, iii) pressing and iv) sintering under Ar / 4 % H{sub 2} + H{sub 2}O gas mixture. In this paper, triuranium octoxide (U{sub 3}O{sub 8}) is used as an additional raw powder in the LWR MOX fuel manufacturing. One of the objectives of this work is to understand the sintering behavior of U{sub 3}O{sub 8} under Ar / 4 % H{sub 2} + H{sub 2}O gas mixture and its effect when added to a UO{sub 2}-PuO{sub 2} mixture. A special attention is given on U{sub 3}O{sub 8} reduction, phase transitions, redox reactions and oxygen partial pressure (pO{sub 2}) evolution during shrinkage: the impact of these phenomena on the densification kinetics is discussed, especially the de-densification phenomenon which could occur at high temperature. For that purpose a specific dilatometer, coupled with a zirconia probe implemented at furnace inlet and outlet has been used which allows following continuously and simultaneously the fuel shrinkage and the oxygen partial pressure variation as a function of temperature. The experiments have shown that, contrary to UO{sub 2}, the sintering of milled PuO{sub 2} is accompanied by a de-densification at high temperature. This de-densification phenomenon already occurs with co-milled 71 at% UO{sub 2} + 29 at% PuO{sub 2}, but with a smaller magnitude. When pure milled U{sub 3}O{sub 8} is sintered, no de-densification phenomenon is observed but the low density of the sintered pellet could hide such a phenomenon. The incorporation of U{sub 3}O{sub 8} into a UO{sub 2}-PuO{sub 2} master blend containing 29 at% of plutonium leads to an exacerbation of the de-densification phenomenon with a higher loss on ignition suggesting that U{sub 3}O{sub 8} contributes also to the de-densification phenomenon. This could be due to high temperature reactions involving the initial carbon content of the raw U{sub 3}O{sub 8} powder. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Synroc Waste-forms for Immobilization of Advanced Reprocessing Wastes - 16167

EURO-GANEX, an advanced reprocessing technology, aims to separate both major (U, Pu) and minor (Np, Am, and Cm) actinides together from fission products. As the final waste composition is free from actinides - as opposed to current process waste, e.g. from PUREX - adapted immobilization matrices need to be developed. Synroc is a potential wasteform that has proven itself to be efficient to immobilize high-level nuclear wastes (HLW). In this study, a new composition of Synroc, 'Synroc-Z', is designed and characterized. The primary modification is to reduce the amount of zirconolite phase, which acts as the main host for actinides. The optimum hot-press conditions for Synroc-Z samples with a waste loading of 20 wt% were determined to be: 1200 deg. C and 20 MPa for 3 h. The durability of Synroc-Z was examined via the MCC-1 test at 90 deg. C. It shows similar leach resistance to Synroc-C, which is much better than current glass waste-forms. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Analysis of Legacy {sup 85}Kr Waste Form Samples - 16319

In the late 1970's, the Idaho Chemical Processing Plant conducted an R and D program on krypton encapsulation. Legacy samples composed of {sup 85}Kr incorporated in solid zeolite 5A material, and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing, have been preserved. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. After receipt and repackaging, the samples were examined by x-ray imaging and x-ray tomography. Two of the capsules appear to contain an amorphous mass within the capsules. An image of one capsule clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. An image of another capsule shows similar bead-like material within the intact capsule. One of the capsules was previously opened. The end of this capsule appears to have been cut off, and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of this capsule. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. Based on the initial nondestructive assay (NDA) and the fact that there are two obviously breached samples, a plan was developed for exploratory tests that would be conducted with the breached specimens before opening the three intact capsules. Preparations are being made to determine the fraction of krypton/xenon remaining in the waste form matrix and the amount of rubidium remaining in the waste form matrix. The inner surface of the breached capsules will be examined for corrosion. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

The DISTINCTIVE University Consortium: Transforming High Activity Materials Research in the UK - 18212

The DISTINCTIVE consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) is a partnership of 11 research universities, working to decommission and clean up the legacy of civil nuclear energy generation, sponsored by the Engineering and Physical Sciences Research Council (EPSRC). An important component of the research programme is to develop the skills and facilities to address key research questions, requiring the use of highly radioactive materials. Here we highlight the outcomes of four research projects in which access to state of the art research facilities for working with radioactive materials, supported by international partnerships, has advanced research in plutonium storage and immobilisation, and the disposal of spent nuclear fuels. First, we have developed a viable formulation and process for immobilisation of the UK plutonium stockpile, including commissioning of a radiological hot isostatic press and validation studies using plutonium-239. Second, we have further underpinned the interim storage of PuO{sub 2}, through understanding of water and chlorine sorption / desorption, pertinent to the historic practice of containment within, and degradation of PVC bags. Third, we have developed a novel methodology to investigate the in situ radiolysis induced dissolution of actinide oxide thin films, using synchrotron radiation, to underpin the geological disposal of spent nuclear fuel. Fourth we monitored in situ the oxidation and self-ignition mechanism of uranium carbide, a legacy fuel from the Dounreay Prototype Fast Reactor that is waiting for oxidation treatment prior to permanent storage and/or disposal. Through this programme of research, the DISTINCTIVE consortium is rebuilding the UK capability for research with transuranic materials, which, hitherto, was eroded over decades of disinvestment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Monolithic U-10Mo Fuel Development for U.S. High Performance Research Reactors

To support the U.S Department of Energy (DOE), National Nuclear Security Administration (NNSA) mission to reduce the threat posed by the civilian use of highly enriched uranium (HEU) worldwide, the US High Performance Research Reactor (US HPRR) fuel development program is pursuing development and qualification of a new high-density monolithic U-Mo fuel to facilitate conversion of five higher power research reactors located in the U.S. (ATR, HFIR, NBSR, MIT and MURR) and one critical facility (ATR-C) from HEU to low enriched uranium (LEU). These reactors require higher uranium fuel density than is achievable with dispersion fuel systems used to convert lower power research reactors. The down selected fuel system consists of U-10Mo alloy fuel foils having a thin Zr diffusion barrier interlayer, clad in 6061 Al alloy by hot isostatic pressing. In order to support fabrication development, fuel performance evaluations, generic fuel qualification through the Nuclear Regulatory Commission (NRC), and eventual reactor licensing, a series of irradiation experiments and testing campaigns are being planned. The tests are designed to provide data needed for fabrication process down selection, to establish acceptable fuel performance under normal operating conditions and anticipated transients, and to demonstrate successful scale-up to prototypic reactor-specific fuel element designs. A comprehensive testing program is being executed as part of the US HPRR fuel development program to develop and qualify a new monolithic UMo fuel system to allow conversion of five high power research reactors and one critical facility within the US in support of the NNSA mission to convert research reactors worldwide from HEU to LEU fuels. A significant outcome of this program will be the qualification and eventual licensing of a new plate-type fuel system for use in research reactors through the NRC for the first time in nearly 3 decades. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Synroc Waste Treatment Facility for Fission-based Molybdenum-99 Production - 19335

ANSTO is building a nuclear waste treatment plant that represents a significant step change in the treatment of intermediate-level waste (ILW) from Molybdenum-99 nuclear medicine production. This industrial scale facility shall deploy ANSTO Synroc technologies; a wasteform tailored for immobilizing fission products, an integrated process line for converting liquid waste into free-flowing powder, automated canister filling and sealing, and a hot isostatic pressing (HIP) system designed for remote operation. The HIP represents the last processing step to consolidate the wasteform into a final durable solid compact with the required performance characteristics. This paper shall outline the various phases of development in realising the final plant design as well as timelines to progress through cold- and hot-commissioning. Further, plans for storage and final disposal of the HIPed canister product will be discussed. (authors)

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

Creep-Fatigue Properties of Additional 316H PM-HIP Materials Fabricated from Different Powder Compositions and Processing Routes

The process of powder metallurgy (PM) hot isostatic pressing (HIP) works by consolidating powdered materials at relatively high temperature and pressure to form near-net-shaped components. Ideally, PM-HIP production methods can reduce component lead time and improve designs for high-temperature reactors and/or microreactors. To introduce PM-HIP into Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, it is necessary to show adequate material properties regarding creep, high-temperature low-cycle fatigue, and creep fatigue. However, prior work has shown that the creep-fatigue cycles to failure for PM-HIP 316H stainless steel are greatly reduced compared to the conventional, wrought product. This work continued creep-fatigue analysis on a 316H stainless steel with lower oxygen and nitrogen contents and at different HIP parameters than previously analyzed. The objective was to better understand what is causing the reduced PM-HIP 316H performance so improvements can be made PM-HIP 316H creep-fatigue lifetimes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Determining the source characteristics of explosions near the Earth's surface

We present a method to determine the source characteristics of explosions near the airearth interface. The technique is an extension of the regional amplitude envelope method and now accounts for the reduction of seismic amplitudes as the depth of the explosion approaches the free surface and less energy is coupled into the ground. We first apply the method to the Humming Roadrunner series of shallow explosions in New Mexico where the yields and depths are known. From these tests, we find an appreciation of knowing the material properties for both source coupling/excitation and the free surface effect. Although there is the expected tradeoff between depth and yield due to coupling effects, the estimated yields are generally close to the known values when the depth is constrained to the free surface. We then apply the method to a regionally recorded explosion in Syria. We estimate an explosive yield less than the 60 tons claimed by sources in the open press. The modifications to the method allow us to apply the technique to new classes of events, but we will need a better understanding of explosion source models and properties of additional geologic materials.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2016

The tropical rain belts with an annual cycle and a continent model intercomparison project: TRACMIP

This paper introduces the Tropical Rain belts with an Annual cycle and a Continent Model Intercomparison Project (TRACMIP). TRACMIP studies the dynamics of tropical rain belts and their response to past and future radiative forcings through simulations with 13 comprehensive and one simplified atmosphere models coupled to a slab ocean and driven by seasonally-varying insolation. Five idealized experiments, two with an aquaplanet setup and three with a setup with an idealized tropical continent, fill the space between prescribed-SST aquaplanet simulations and realistic simulations provided by CMIP5/6. The simulations reproduce key features of the present-day climate and expected future climate change, including an annual-mean intertropical convergence zone (ITCZ) that is located north of the equator and Hadley cells and eddy-driven jets that are similar to the present-day climate. Quadrupling CO 2 leads to a northward ITCZ shift and preferential warming in Northern high-latitudes. The simulations show interesting CO 2 -induced changes in the seasonal excursion of the ITCZ and indicate a possible state-dependence of climate sensitivity. The inclusion of an idealized continent modulates both the control climate and the response to increased CO 2 ; for example it reduces the northward ITCZ shift associated with warming and, in some models, climate sensitivity. In response to eccentricity-driven seasonal insolation changes, seasonal changes in oceanic rainfall are best characterized as a meridional dipole, while seasonal continental rainfall changes tend to be symmetric about the equator. Finally, this survey illustrates TRACMIP’s potential to engender a deeper understanding of global and regional climate phenomena and to address pressing questions on past and future climate change.

54 ENVIRONMENTAL SCIENCES↗