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At least 775 records · Page 43

FFTF HT9 Cladding Microstructure Characterization

The sodium-cooled fast reactor (SFR) is a promising candidate for next generation nuclear reactors, operating at extreme conditions which include high temperatures (>500?C core outlet temperature) and significant neutron damage. High-Cr martensitic HT9 steel is an excellent candidate for SFR cladding and duct material due to its compatibility with liquid sodium, good thermal conductivity, resistance to void swelling, and strong creep rupture strength [1-4].However, the harsh in-core environment of SFRs can cause complex microstructural changes and mechanical property degradation in HT-9. Ensuring the safe use of HT9 cladding for metallic fuel requires both a thorough understanding of its mechanical response to microstructure evolution as well as reliable microstructure-sensitive modeling predictions. Microstructure-sensitive modeling of high temperature creep behavior in HT9 cladding for SFR applications currently lack experimental data to model the phenomena accurately. To fill this need, methods to perform microstructural characterization have been developed and performed on HT9.

36 MATERIALS SCIENCE↗

Microstructure and microchemistry changes at U-10Mo fuel/AA6061 cladding interfaces with varying hot isostatic pressing conditions

Monolithic uranium - 10 wt.% molybdenum (U-10Mo) is a promising high-assay low-enriched uranium fuel system for nuclear reactors used in research, medical isotope production, and remote power applications. During manufacturing, AA6061 cladding is bonded to the U-10Mo fuel plate via hot isostatic pressing, during which diffusion and fuel/cladding chemical interaction occurs at plate edges. Furthermore, the microstructure and microchemistry changes that occur at fuel/cladding interfaces are important to understand as a function of process parameters to develop a reliable fuel fabrication process and to meet the desired specifications. Here, microstructural and microchemical changes are studied using complementary electron microscopy and atom probe tomography for two manufactured fuel plates with varied hot isostatic pressing (HIP) parameters. Results highlight that modifying thermomechanical processing parameters significantly changed the interaction layer thickness between U-10Mo and AA6061 by an order of magnitude. In addition to this the distribution and concentration of elements (i.e., Al, Si, Mg) from cladding to fuel was also investigated.

36 MATERIALS SCIENCE↗

Hollow-core revolver fibre with a double-capillary reflective cladding

We report the fabrication of the first hollow-core revolver fibre with a core diameter as small as 25 μm and an optical loss no higher than 75 dB km{sup -1} at a wavelength of 1850 nm. The decrease in core diameter, with no significant increase in optical loss, is due to the use of double nested capillaries in the reflective cladding design. A number of technical problems pertaining to the fabrication of such fibres are resolved. (fiber optics)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

GTRF Cladding Wear Modeling and Experimental Validation

Coolant flow-induced relative motions between certain structural components in the cores of lightwater nuclear reactor cores, a phenomenon called Grid-To-Rod Fretting (GTRF), can result in progressive wear damage. Wear-through of the cladding on fuel rods can introduce radioactive species into the coolant loop. Such an event is termed a “leaker” and may cause premature and expensive reactor shut-down. GTRF results in wear scars of various sizes, shapes, and depths on the sides of the fuel rod cladding. Fretting damage may also occur on the springs and dimples that hold the fuel rods in place, resulting in the formation of gaps. Due to variability in the operating conditions of the reactor (burn-up rates and operating campaigns) and the associated variable cladding surface oxide film growth and removal rates, the wear coefficient is unlikely to remain constant over time. Therefore, constant wear rate models do not accurately account for transients in the operating conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of LWR fuels with enhanced accident tolerance. Final report

After a tsunami triggered plant black out at the Fukushima Daiichi site, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2 – zirconium alloy fuel system. This accident tolerant fuel (ATF) alternative should better tolerate a loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric, Oak Ridge National Laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT, C26M or FeCrAl ODS, which has been named IronClad by GE. Extensive testing and evaluation has been conducted to determine the suitability of FeCrAl for the entire fuel cycle from rod fabrication to used fuel disposition. Most of the environmental characterization was conducted to evaluate the behavior of IronClad under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher parasitic neutron absorption and potential tritium release into the coolant effects can be minimized by design and operational measures. IronClad material can be manufactured using traditional melting and forging (C26M) or powder metallurgy processes (APMT & FeCrAl ODS). Tubes can be made to full length of 5 m with a wall thickness of 0.4 mm by pilgering or drawing. The fuel rods can be readily hermetically sealed in an industrial production setting by welding end caps to the thin wall tubing using either TIG or resistance welding. The implementation of FeCrAl IronClad cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Site-specific nanoscale characterization of zirconium hydrides in the hydride rim structure of hydrogen-charged zircaloy-4 cladding

Zircaloy-4 cladding tubes were hydrogen-charged using a static pressure hydrogen charging method to experimentally produce a zirconium-hydride rim structure similar to that observed in Zircaloy-4 cladding tubes after operation in commercial light-water nuclear reactors. Detailed characterization was performed using electron energy loss spectroscopy (EELS) and four-dimensional scanning transmission electron microscopy (4D-STEM). By mapping the position of the plasmon peak in the low-loss EELS spectrum, it was determined that the zirconium hydrides formed were predominantly solid δ-phase zirconium hydrides. 4D-STEM was used to generate phase and strain maps at nanoscale. The interactions between two zirconium hydride platelets less than 200 nm apart leads to the localized lattice rotations and possible phase change, revealing a potential hydride growth mechanism.

4D-STEM↗

Multidimensional Multiphysics Simulation of Nuclear Fuel Behavior

Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non- axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our multidimensional, multiphysics approach to analyze a missing pellet surface problem. The next example is the analysis of cesium diffusion in a TRISO fuel particle with defects. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Advanced Nuclear Fuels and Materials Development and Philosophy of the DOE Advanced Fuels Campaign

The mission of the Advanced Fuels Campaign (AFC) is to perform Research, Development, and Demonstration (RD and D) activities for advanced fuel forms (including cladding) to enhance the performance and safety of the nation's current and future reactors; enhance proliferation resistance of nuclear fuel; effectively utilize nuclear energy resources; and address longer-term waste management challenges. This includes development of a state-of-the art Research and Development (R and D) infrastructure to support the use of a 'goal-oriented science-based approach.' AFC uses this approach to develop a fundamental understanding of fuel and cladding fabrication methods and performance under irradiation, enabling the pursuit of multiple fuel forms for future fuel cycle options. This approach includes fundamental experiments, theory, and advanced modeling and simulation. One of the most challenging aspects of AFC is the management, integration, and coordination of major R and D activities across multiple organizations. AFC interfaces and collaborates with other Fuel Cycle Technologies (FCT) campaigns, universities, industry, various DOE programs and laboratories, federal agencies (e.g., Nuclear Regulatory Commission [NRC]), and international organizations. Key challenges are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. AFC is challenged with the research and development of fuels for two different reactor technology platforms, namely transmutation fuel for non-water cooled reactors and advanced fuel development for light water reactor (LWR) systems. This presentation will provide an overview of the Advanced Fuel Campaign activities including the near term activities in advanced nuclear fuel system research and development as well as highlighting general activities pursued to improve the study, modeling, and simulation of irradiated nuclear materials. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Chemical Dissolution of Simulant FCA Cladding and Plates

The Savannah River Site (SRS) has received some fast critical assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) for disposition. Among the JAEA FCA fuel are approximately 7090 rectangular Stainless Steel clad fuel elements. Each element has an internal Pu-10.6Al alloy metal wafer. The thickness of each element is either 1/16 inch or 1/32 inch. The dimensions of each element ranges from 2 inches x 1 inch to 2 inches x 4 inches. This report discusses the potential chemical dissolution of the FCA clad material or stainless steel. This technology uses nitric acid-potassium fluoride (HNO 3 -KF) flowsheets of H-Canyon to dissolve the FCA elements from a rack of materials. Historically, dissolution flowsheets have aimed to maximize Pu dissolution rates while minimizing stainless steel dissolution (corrosion) rates. Because the FCA cladding is made of stainless steel, this work sought to accelerate stainless steel dissolution.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Using Falcon to develop an RIA PCMI failure criteria

This paper documents the development of failure criteria for both pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rod failure due to pellet cladding mechanical interaction (PCMI) during a reactivity insertion accident (RIA). This study was initiated to use cladding mechanical properties using data from separate effects tests to provide a basis to translate nuclear safety research reactor (NSRR) test data to commercial light water reactor conditions. The approach employed for the revised licensing criteria combined elements of experimental results and analytical evaluations to develop a fundamental understanding of fuel behavior during RIA events. The transient behavior of intermediate and high burnup fuel rods using well characterized RIA simulation tests was established. Cladding mechanical properties using data from separate effects tests was also developed. The Zircaloy cladding mechanical properties database of separate effects tests was used to develop the cladding integrity models. The cladding integrity models were then used in combination with the application of the Falcon fuel behavior code to translate NSRR test data to commercial light water reactor conditions. Because of the complex manner in which burnup influences fuel rod failure, it was not possible to develop a fuel rod failure threshold directly from the experimental data. An alternative approach was required that combined analytical modeling and experimental data to develop the PCMI-related contribution to the failure threshold. The analytical approach computed the radial average peak fuel enthalpy required to cause cladding failure by PCMI as a function of hydrogen content. This was accomplished by applying a critical strain energy density (CSED)-based cladding integrity model using data from total and uniform elongation tests of irradiated low-tin Zircaloy-4, ZIRLO, and Zircaloy-2 cladding materials. The CSED was developed as a function of cladding hydrogen content because hydrogen induced embrittlement is a primary effect of burnup. It was found that the failure threshold decreased with hydrogen content due to the increase in PCMI loading as a result of gap closure effects and by the decrease in cladding ductility with hydrogen accumulation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Ageing Management for Extended Long-Term Dry Storage of Spent Nuclear Fuel and Transportation

The principal objectives of the Coordinated Research Project (CRP) and CRADA were to (1) investigate how ageing effects leading to degradation of materials used in the spent fuel dry storage systems could be managed by ageing management programs (AMPs) and (2) using existing AMPs as a basis, establish guidance on how to develop, generate, and maintain AMPs for dry storage systems of spent nuclear fuel (SNF) that can be accomplished in various ways. The PI (Dr. Liu) is an internationally recognized expert on ageing management for license renewal of nuclear power plants and independent spent fuel dry storage installations. Dr. Liu was invited by IAEA to participate in this CRP and served as Chair of the Working Group, supporting the IAEA CRP Lead, with other members from Argentina, the Czech Republic, France, Germany, Hungary, Japan, Pakistan, Spain, Switzerland, United Kingdom, and the United States of America. Argonne’s scope of work in the CRADA for the CRP included (1) ageing management guidance documents developed by Argonne for DOE and used by the Nuclear Regulatory Commission (NRC) and industry and (2) research leveraged from the DOE Office of Nuclear Energy on the mechanical properties of high-burnup fuel cladding and the ARG-US remote monitoring systems technology developed for the DOE Packaging Certification Program, Office of Packaging and Transportation, Office of Environmental Management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Dynamic System Scaling Applied to Zr-4 Cladding under RIA Conditions

The research and development lifecycle for new fuel designs currently has an ideal 20-25 year timeline. Efforts are currrently underway to accelerate fuel design and development using high performance fuel modeling codes and reduced scale experiments with new instrumentation techniques. However current tools available require a full scale prototype to be designed and tested for fuel qualification. The thermal-hydraulics community has leveraged scaling methodologies to use reduced scale integral effect test facilities (Westinghouse AP600/1000, NuScale Power Module) in lieu of building full scale prototypes for reactor design certification. Dynamic System Scaling (DSS) is a recent development to scaling methodologies that enable the tracking and quantification of dynamic scaling distortions experienced throughout a transient. Recent TREAT experiments performed reactivity initiated accident (RIA) pulses on UO2 fuel rodlets in Zr-4 cladding and were accompanied by BISON predictions for the cladding temperature. DSS is applied to pyrometer data of outer cladding temperatures and two BISON predictions after peak cladding temperature has been reached. One BISON case uses a constant gap conductance value between the fuel and cladding and the other using a conventional gap conductance model. Results from the DSS analysis show that there is good agreement between both BISON predictions and the experimental data and that there is not a significant distortion seen between using either gap conductance model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Advancements in modeling fuel pulverization and cladding behavior during a LOCA

During a loss of coolant accident (LOCA), there is the possibility of nuclear fuel rods to undergo a three step process known as fuel fragmentation, relocation, and dispersal (FFRD). The chance of FFRD occurring increases as the fuel burnup increases. To support the nuclear industry's desire to increase the discharge burnup of nuclear fuels in light water reactors (LWRs) it is imperative to understand the mechanisms driving the evolution of FFRD. In this work, a multiscale modeling modeling approach is used to garner insight into underlying mechanisms leading to the ne fragmentation (also known as pulverization) of nuclear fuel during a LOCA. This report includes a summary of the atomistic and phase-field studies to develop a new pulverization criterion for use in the engineering scale Bison fuel performance code. Details are also provided on cladding modeling improvements related to hydrogen/hydride embrittlement and damage, and anisotropic thermal creep. The new models are used on the existing integral and separate effects LOCA validation cases available in Bison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

FRAPCON analysis of cladding performance during dry storage operations

There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

Dry Storage↗

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

Houts, Michael G.↗

Overview of the multifaceted activities towards development and deployment of nuclear-grade FeCrAl Alloys

A large effort is underway under the leadership of US DOE Fuel Cycle R&D program to develop advanced FeCrAl alloys as accident tolerant fuel (ATF) cladding to replace Zr-based alloys in light water reactors. The primary motivation is the excellent oxidation resistance of these alloys in high-temperature steam environments right up to their melting point (roughly three orders of magnitude slower oxidation kinetics than zirconium). A multifaceted effort is ongoing to rapidly advance FeCrAl alloys as a mature ATF concept. The activities span the broad spectrum of alloy development, environmental testing (high-temperature high-pressure water and elevated temperature steam), detailed mechanical characterization, material property database development, neutron irradiation, thin tube production, and multiple integral fuel test campaigns. Instead of off-the-shelf commercial alloys that might not prove optimal for the LWR fuel cladding application, a large amount of effort has been placed on the alloy development to identify the most optimum composition and microstructure for this application. The development program is targeting a cladding that offers performance comparable to or better than modern Zr-based alloys under normal operating and off-normal conditions. This paper provides a comprehensive overview of the systematic effort to advance nuclear-grade FeCrAl alloys as an ATF cladding in commercial LWRs.

Field, Kevin G↗

Advancements in modeling fuel pulverization and cladding behavior during a LOCA

During a loss-of-coolant accident (LOCA), it is possible for nuclear fuel rods to undergo a three-step process known as fuel fragmentation, relocation, and dispersal (FFRD). The chance of FFRD occurring increases as the fuel burnup increases. To support the nuclear industry's desire to increase the discharge burnup of nuclear fuels in light-water reactors (LWRs), it is imperative to understand the mechanisms driving the evolution of FFRD. In this work, a multiscale modeling approach is used to garner insight into underlying mechanisms leading to the fine fragmentation (also known as pulverization) of nuclear fuel during a LOCA. This report includes a summary of the atomistic and phase-field studies to develop a new pulverization criterion for use in the engineering-scale Bison fuel performance code. Details are also provided on cladding model improvements related to hydrogen/hydride embrittlement and damage and anisotropic thermal creep. The new models are used on the existing integral and separate effects LOCA validation cases available in Bison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The Development of Reversible Bending Fatigue Tester and Its Application to High Burnup Spent Nuclear Fuel Vibration Integrity Investigation under Normal Transportation

For high burn-up spent nuclear fuel (SNF), it is expected that the spent fuel cladding will have a high population of microcracks and hydrides, including macro-hydrides and micro-hydrides. This will reduce the stress intensity required to advance the crack growth. The linking of these microcracks during vibration loading also reduces the fatigue threshold/incubation period significantly, accelerating fatigue failure. In addition to the damages in cladding, the microstructure of comprising fuel pellets and interfaces of fuel rod have changed dramatically after the high burn-up process. These modifications have a direct impact on the structural integrity and vibration response of SNF rods in transportation.

Fuel Transport↗