Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “TRISO”

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

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 829 records · Page 46
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Fuels and Materials Challenges for the Very High Temperature Reactor

High Temperature Gas-cooled Reactors (HTGRs) featuring particle fuel are a very mature technology. Commercial deployment occurred in the mid-1980's with the Fort St.Vrain and Thorium Hochtemperatur Reaktor by providing electricity to the grids in the United States and West Germany, respectively. The technology was then adopted by Japan and China with the operation of the High Temperature Test Reactor in Oarai, Japan and the High Temperature Reactor (HTR-10) in China. These commercial reactors typically operate with an outlet coolant temperature of 750 deg. C to 800 deg. C. Increasing the outlet temperature of the HTGR to even higher temperatures above 900 deg. C will improve the thermodynamic efficiency of the system and enable application of a new class of gas reactor, the very high temperature reactor (VHTR), to provide process heat, electricity, and hydrogen to a broad range of industrial customers with the attendant benefits of improved energy security and reduced CO{sub 2} emissions. However, the increase in coolant outlet temperature presents a number of technical challenges associated with fuels and materials. For over a decade, the U.S. Department of Energy has been sponsoring a broad program of research and development with a goal of addressing the technical fuels and materials challenges to support deployment of a VHTR that could safely and economically operate at outlet temperatures of 750 to 950 deg. C. In summary, fuels and materials challenges exist with deploying a VHTR with an outlet temperature of 950 deg. C in the areas of structural materials, graphite and TRISO fuel. However, R and D is underway or has recently been completed suggesting that technical solutions are possible. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An Overview of Material Models in the BISON Fuel Performance Code

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and supports the use of one-, two-, and three-dimensional meshes. BISON is designed to be a multi-fuels code. Although primarily applied to Light Water Reactor (LWR) fuel to date, the code has been used to analyze TRISO-coated particle fuel and metal fuel in rod and plate form, design and interpret fuel irradiation experiments and investigate novel fuel concepts. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models which are often complex and nonlinear. Nearly all of the material models currently available in BISON are empirical, using correlations from existing fuel performance codes or material property libraries. Although these models perform adequately under many conditions, they have well-understood limitations. One obvious weakness is that the models are only valid in the domain of the associated experimental data, and extrapolating beyond this domain can result in substantial errors. A further weakness is that these models are often correlated in terms of imprecise quantities, such as burnup. Materials can, for example, reach a given burnup under very different circumstances, leading to very different material behavior. To overcome these weaknesses, efforts are underway with BISON to develop models based on a multi-scale approach, where material behavior is described with a series of interdependent microstructure based models. This paper provides an overview of the material models available in BISON. To keep the scope manageable, only LWR fuel will be considered. The first section focuses on UO{sub 2} fuel, identifying the models generally required and briefly describing those that have been implemented in the code. The second section repeats this approach for Zircaloy cladding. A final section provides examples of areas where multi-scale material modeling has been applied to BISON. BISON is a modern 2D and 3D nuclear fuel performance code that is applicable to multiple fuel forms. A wide variety of complex and nonlinear material models are required to simulate fuel behavior. Focusing on UO{sub 2} fuel and Zircaloy cladding (LWR fuel), the set of empirical models currently implemented in BISON have been reviewed. To overcome some of the deficiencies associated with empirical models, multi-scale material models are being developed. Two multi-scale models that have been used in BISON were reviewed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Recent Advances in Analytical Electron Microscopy for Irradiated Materials

Analytical electron microscopy (AEM) in the scanning transmission electron microscope (STEM) has long been a staple of post-irradiation examination (PIE) and materials development campaigns. Recently, new advances in both STEM hardware and post-processing data-analysis software, along with ancillary advances such as in focused ion beam (FIB) sample preparation, has led to exciting advances in the quantity and quality of data that can be obtained. This talk will summarize recent experiments performed at Oak Ridge National Laboratory (ORNL), especially work on neutron-irradiated materials and fuels taking advantage of a newly-installed latest-generation AEM tool in the ORNL irradiated materials laboratories. ORNL installed a new FEI Talos F200X microscope, a 200 keV, high-brightness Schottky emission AEM system optimized for high-speed, high-efficiency X-ray mapping experiments. The F200X features the FEI 'SuperX' four-sector silicon drift detector (SDD) X-ray detector, which provides 0.9 sr solid angle of collection. By combining the high-brightness electron source, latest-generation electron optics, and fast computers and pulse processors, X-ray mapping speed is increased by a factor of 50-100 x compared to the previous-generation instruments. In addition to improvements in hardware, data mining via multivariate statistical analysis (MVSA) allows for maximum insight to the data obtained from the new hardware. MVSA data mining of the X-ray maps allows, for instance, unbiased computational identification of the phases present, and discovery of small phases or precipitates not expected by the human analyst. This talk will present examples of experiments performed using the most modern AEM tools and software. Silicon carbide is a major candidate for next-generation nuclear applications. Nano-infiltration by transient eutectic (NITE) SiC requires low sintering temperatures. X-ray mapping of the NITE structure find pockets of different oxides, such as Zr-Y-oxide, Y-Al-oxide, and yttria. MVSA data mining finds the regions where different oxides are present and indicates that Y additions preferentially wet grain boundaries, where Zr additions are preferentially concentrated at grain triple points. SiC coatings from AGR-I experiment TRISO particles were found to contain significant quantities of fission products (e.g., Pd) as well as actinides (U, Pu). MVSA data mining found complex precipitates and grain boundary films of the fission products and actinides deep (∼several microns) into the SiC layer. Advanced AEM methods, combined with MVSA data mining, makes possible both rapid and detailed PIE. Ferritic alloys, such as accident-tolerant FeCrAl grades, benefit greatly from advanced AEM examination. Neutron-irradiated FeCrAl precipitates the detrimental Cr-rich α' phase. Because the SuperX SDD system has four detectors, it is possible to obtain high X-ray count rates even on a highly tilted specimen, as might be needed to obtain STEM images of dislocation loops. Results on neutron-irradiated FeCrAl indicates that α' size and number density can be obtained by AEM, and MVSA results determine that minor elements (such as Si) has complex radiation-induced segregation behavior at grain boundaries. We have also developed and validated computer models of the AEM experiment, allowing different microstructures' X-ray map results to be simulated for different STEM instruments. This allows X-ray mapping to be more quantitative, such as by estimating what size nanoclusters in an oxide-dispersion strengthened steel would be visible or invisible in STEM. In summary, new advanced AEM technology provides a major leap forward in experiments possible in PIE of irradiated materials. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Positron Annihilation Spectroscopy Investigation of Vacancy Clusters in Neutron-Irradiated 3C-SiC

Silicon carbide (SiC) has been widely applied or considered in nuclear systems as fuel and structural materials since 1960's, given its high-temperature stability, chemical inertness, and exceptional irradiation resistance. Chemically vapor deposited (CVD) SiC has been considered as the pressure vessel materials for the TRISO fuel particles used in high temperature gas-cooled reactors or novel accident tolerant fuel concepts for light water reactors. The use of SiC/SiC composites as primary structural materials has also been proposed in fusion reactors. Driven by the need to enlarge the safety margins of nuclear reactors in accident scenarios, R and D of accident-tolerant fuel has become an important topic in the nuclear engineering and materials community. A continuous fiber SiC/SiC composite is under consideration for Zr alloy cladding replacement. Microstructural development and mechanical property change in SiC subject to neutron and ion irradiation have been studied extensively and understood well. However, the information of small vacancy cluster (i.e., those invisible in TEM) evolution in SiC exposed to neutron irradiation is still lacking, and such small clusters is suspected to play an important role in swelling, radiation creep, and thermal conductivity degradation. In this study, positron annihilation spectroscopy is used to characterize the vacancy defects in 3C-SiC exposed to various neutron irradiation conditions to capture the dependence of temperature and irradiation dose. The obtained vacancy cluster distribution is then linked to the measured swelling behavior to provide insight into the relationship between microstructure and physical property of SiC. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

A Preliminary Overview of Metal Fuel Models in BISON

BISON is a modern nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. It is a finite element-based code capable of solving the energy, mechanics, and species diffusion equations in a fully-coupled sense. BISON also utilizes thermal and mechanical contact algorithms and variety of nonlinear material models. BISON is applicable to steady and transient fuel behavior and supports the use of one-, two-, and three-dimensional meshes. Although thus far used primarily to simulate light water reactor fuel, BISON is designed to be a multi-fuels code. As such, BISON has been used to analyze TRISO-coated particle fuel and metal fuel in rod and plate form, design and interpret fuel irradiation experiments and investigate novel fuel concepts. This publication provides a preliminary list of metal fuel-related material models recently implemented in BISON. Some preliminary results from a simulation of the EBR-II X447 experiment are shown, which begins to demonstrate how all these models work together. A follow-up publication will be presented with more detail at the ANS meeting in New Orleans during the embedded topical meeting titled 'Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors'. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Corrosion of Candidate Materials in Molten FLiBe Salt for Fluoride-Salt-Cooled High-Temperature Reactor (FHR)

Molten fluoride salts have attracted considerable attention because of their desirable thermophysical properties that lead to superior heat transfer. Recently, fluoride salts are being considered as coolants in the fluoride salt-cooled high temperature reactors (FHRs). The successes of the molten salt reactor experiment (MSRE) program in 1950's-70's at Oak Ridge National Laboratory (ORNL) has also driven the pursuit of molten salt nuclear reactor as one of the lead next generation nuclear reactor concepts. Unlike MSRE where the uranium fuel was dissolved in the molten fluoride salt, the FHR concept is expected to use molten salt (2LiF-BeF{sub 2}, FLiBe) only as coolant in conjunction with solid TRISO fuel particles immersed in the molten salt. One of the important challenges in FHR development is corrosion of structural materials in high-temperature molten 2LiF-BeF{sub 2} (FLiBe) salt. Currently, 316 stainless steel (nominal composition: Fe-12Ni-17Cr-2.5Mo-1Si-2Mn-0.03C) and Hastelloy N (nominal composition: Ni-7Cr-16Mo-5Fe-1Si-0.8Mn) are deemed as promising candidate alloys for the construction of some of the structural components of FHR. 316 stainless steel is ASME code certified for high temperatures (up to 760 deg. C) applications. Hastelloy N has been shown to have outstanding corrosion resistance in molten FLiBe but present code certification data limit the use of this alloy to only about 704 deg. C. Modifications of Hastelloy N with higher creep strength are presently being developed at ORNL. Other materials such graphite, SiC/SiCf and C/Cf composites, are also considered for structural applications in the FHR. The presence of these multiple materials in a molten FLiBe salt medium can further exacerbate and complicate corrosion effects. Therefore, it is necessary to understand the corrosion behavior of these materials in molten FLiBe salt. The corrosion behavior of Hastelloy N, 316 stainless steel, and CVD SiC and two types of SiC-SiC{sub f} composites have been tested in molten FLiBe salt at 700 deg. C. When tested in metal-lined crucibles both Hastelloy N and 316 stainless steel showed very good corrosion resistance. Samples tested in graphite crucibles showed greater propensity for corrosion and carburization was observed in the near surface regions of the alloys. CVD SiC showed outstanding corrosion resistance. However, the two SiC-SiC{sub f} composites exhibited corrosion attack implying that the manufacturing process may play a role in the corrosion of these composites. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Property evaluation and microstructure characterization of the A3-3 matrix graphite - 18440

Spherical fuel elements with TRISO fuel will be used in the pebble-bed high temperature gas-cooled nuclear reactor. The fuel will be located within the center region of the elements, and will be suspended in, and surrounded by a shell of matrix graphite. The matrix graphite provides structural support to the fuel, prevents fuel damage, and allows for heat transfer from the fuel to the coolant. The raw materials for the matrix graphite, type A3-3, includes 64 wt% natural flake graphite, 16 wt% artificial graphite and 20 wt% phenol resin binder. After the final heat treatment, the resulting A3-3 matrix is composed of 71 wt% natural graphite, 18 wt% artificial graphite, and 11 wt% phenol resin-derived coke carbon. The microstructure of A3-3 matrix graphite has direct influences on the material properties, which are crucial to maintaining the integrity and safety of the fuel elements. This presentation will discuss microstructure features, such as lattice, and crystal parameters, porosity, pore size and distribution, and anisotropy. The multiple physical, mechanical and thermal properties of this material, including: elastic modulus, mechanical strength, and oxidation property will also be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment - 18579

The PARFUME (Particle Fuel Model) code was used to predict silver release from tri-structural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO{sub 2}) kernels. Post-irradiation examination (PIE) measurements were performed to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO{sub 2} compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Mechanistic source terms for the XE-100 reactor XSTERM Code Development Approach - 18587

X-Energy is developing the Xe-100 pebble bed modular High Temperature Gas Reactor (HTGR) as an advanced nuclear power source. To achieve its safety objectives, the Xe-100 reactor relies on materials inherent characteristics and passive safety features. Key features include the inherent high temperature characteristics of TRISO-coated fuel particles, graphite moderator, and helium coolant, along with passive heat removal capability of a low-power-density core. This paper will describe the approach adopted by X-Energy to develop and validate mechanistic source terms for the Xe-100. The Mechanistic Source Term (MST) approach for the Xe-100 reactor will build upon that of the prior DOE Next Generation Nuclear Program (NGNP) (Refs. 1, 2, 3 and 4). MST development is closely related to Fuel Qualification (FQ), Licensing Basis Event (LBE) selection and Probabilistic Risk Assessment (PRA). This paper provides a brief overview of the MST road map process. Software plays a crucial role in establishing and quantifying MSTs required for licensing the Xe-100. An insight into X-Energy s comprehensive, integrated software development program for calculating time-dependent radionuclide releases from the Xe-100 over a wide range of events is provided. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation - 18594

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation and evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Preliminary study on applying the method of characteristics to pebble-bed HTR - 18637

Neutron transport in high temperature pebble-bed gas-cooled reactor, which is a promising generation IV reactor candidate with cylindrical core, graphite moderator and spherical TRISO coated particles fuel, is of great interest. However, the stochastic method namely Monte Carlo faces with tremendous simulation of particles and the conventional two-step method featured by homogenization only can give the scalar flux. In this paper, method of characteristics is trying to be applied to pebble bed reactor because of its advantages in arbitrary geometry and the ability to obtain solution in neutronically large domains. Some challenges occur when it comes to HTR. First, the amount of calculation and the memory requirement are huge due to the large quantities of spherical fuel and stochastic distribution of pebbles. Second, the arrangement of characteristics and ray tracing process is more complex comparing to LWR with cuboid geometry because the stochastic distribution of the spherical assemblies and the longer neutron mean free path in pebble bed HTR. Third, the treatment of the multiscale sphere fuel need be detail studied to determine the subdivision of flat source region. At present, preliminary research on this subject is undergoing. This paper is concentrated on the modeling of the pebble-bed reactor core by constructive solid geometry (CSG) which allows complex spatial models to be built using boolean operations, characteristics arrangement and the treatment of boundary condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

AREVA Steam Cycle High Temperature Gas-Cooled Reactor Development and Deployment Plan - 18684

The development and deployment timeline for the AREVA SC-HTGR is among the shortest of all Generation IV reactor technologies. Once demonstrated, the SC-HTGR enters an expanding marketplace beyond electricity generation. This enables industrial growth that is today solely captive to fossil fuels, primarily natural gas supply. The SC-HTGR plant will be a hedge that insulates a wide range of process industries from fossil fuel price and supply volatility in coming decades when the SC-HTGR plant design is ready for deployment. Similar to existing light water reactors, the SC-HTGR is largely immune to fuel price swings for its expected 60 year operating life cycle. Approximately 70% of the SCHTGR plant energy cost is driven by the initial capital investment with fuel being less than 20%. This is entirely the opposite of natural gas fueled power plants where about 70% of the energy cost is directly tied to the cost and availability of fuel and the potential volatility this brings with it over a plant life cycle that could span several decades. This paper details the development and deployment strategy envisioned by AREVA for the SC-HTGR plant. The development phase includes completion of the full scale reactor module design and licensing, build and demonstrate the first commercial scale module, get design certification, and launch the commercial fleet deployment. The development plan takes full advantage of a) a large body of successful pre-application licensing interactions with the NRC under the DOE NGNP project, b) ongoing experiments and test results accumulated through the TRISO particle fuel qualification and nuclear grade graphite characterization R and D activities conducted in the U.S. and internationally, and c) operation of several small scale technology demonstration facilities and large scale commercial design projects. The abundance of the historic HTGR reactor technology demonstration projects provides a high degree of confidence that sufficient technical information is available to begin FOAK plant engineering and licensing followed by commercial demonstration of the first reactor module within a next decade. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Electrochemical Property Determination of Fission Product Elements in Molten Fluoride Salt

The fluoride salt-cooled high temperature reactor (FHR) is a developing nuclear reactor concept that combines many attractive features of previous reactor designs. The FHR concept is also a technological bridge between today's light water reactors (LWRs) which use solid fuel and water coolant, to future technologies that utilize liquid salt coolant and dissolved liquid fuel. The FHR concept incorporates low-pressure liquid fluoride salt, tri-structural isotropic (TRISO) coated particle fuel, and an entirely passive decay heat removal system. FHRs operating at very high temperatures increase the thermal energy transfer and subsequently thermal efficiency. Similar to LWRs, the FHR will feature a cleanup system that performs online processing of the primary coolant in order to mitigate contamination from leaking fuel and moisture. Due to its ionic liquid properties, molten salts can be controlled with various applications of electrochemical properties. Oak Ridge National Laboratory (ORNL) studied these applications greatly during the Molten Salt Reactor Experiment (MSRE) and the Molten Salt Breeder Reactor (MSBR) concept during the 1960's and 70's; however, many technology gaps exist in terms of online salt processing for the FHR. The purpose of this research is to identify the electrochemical properties of leaking fission products as well as contamination from moisture in the system. These properties have been thoroughly studied for molten chloride salts but not as much data is reported for behavior in fluoride salts. Before a cleanup system can be proposed for the FHR concept, these properties will need to be extensively investigated to generate a reliable source of electrochemical data. These results are gathered through an electrochemical system which probes the molten salt at high temperatures (up to 750 deg. C). (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutronics Code Benchmark for Fluoride-Salt-Cooled Reactors

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor design that aims to improve on the efficiency, cost, safety and non-proliferation security of nuclear energy, while exploiting existing technologies to minimize the design and licensing burden. FHRs share many design features with the high-temperature gas-cooled reactor (HTGR), namely using the same fuel form: a bed of graphite spheres, each with thousands of small coated fuel particles (TRISO) dispersed throughout it. However, rather than using a gas as a coolant, the FHR uses a molten salt which changes the behavior of the system; in particular, from a neutronics perspective, the coolant is no longer effectively transparent to neutrons, but rather contributes both to neutron moderation and absorption. It is difficult to perform exploratory reactor physics experiments for new reactor concepts, as the regulation and cost surrounding a fuel experiment is often highly restrictive and prohibitive. As such, researchers and design engineers must rely heavily on simulation capability to develop a working knowledge for how the reactor physics of a pre-prototypical reactor might behave. It is then essential to have confidence that simulation codes provide credible results. As there are no instances of FHRs deployed to date, nor are there any experimental data available on neutron behavior of pebble fuel immersed in salt, the first step is to perform a code-to-code verification of idealized scenarios that capture the physics involved with an FHR system. This summary describes the development of a code-to-code benchmark for FHR with the purpose of code verification and presents preliminary results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Benchmark Study of Infinite Lattice of Pebble Fuel within FLiBe Coolant for TMSR-SF1 Design

Solid Fueled Thorium Molten Salt Reactor (TMSR-SF1) is Chinese Fluoride-salt-cooled High Temperature Reactor (FHR) designed by the TMSR group. The construction of this test reactor is expected in the near years. Using TRISO particle based pebble fuel leads to various challenges in neutronics modeling, such as double heterogeneity effect, preserving exact number of particles, their random distributions in pebble, and etc. For the purpose of verifying the calculations, benchmark study of infinite lattice of pebble fuel within FLiBe coolant for the TMSR-SF1 design has been performed. Back-to-back comparisons have been made between TMSR and MIT researchers using different Monte-Carlo based codes. Very good agreements have been found and the results quantitatively evaluate the neutronic effect of various modeling approaches. In addition, sensitivity study has been carried out for the reference pebble-bed fuel lattice. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Mixed effects associated with air ingress in graphite channels

The next generation of nuclear power plants, such as Modular High Temperature Gas-Cooled Reactors (MHTGR), have passive safety design features to overcome transients during off-normal operation but it is unproven whether their safety systems will perform as per design requirements. The design and construction of a facility, to investigate the passive safety response of the complete reactor system will require significant resources. Therefore, a practical approach is to improve the understanding of this multi-physics or mixed effect problem in laboratory scale set-ups and computational models. The HTGR related separate effects experiments at academic institutions and other laboratories have been developed to provide validation data and complement computational efforts to identify critical issues related to passive safety design. There are several separate effect experimental studies currently under investigation or recently conducted by several academic institutions in U.S and throughout the world. The High Temperature Test Facility (HTTF) has been constructed at Oregon State University to conduct integral thermal-hydraulics experiments for the development of HTGRs. For Depressurized Conduction Cooling (DCC) experiments in HTTF, a mixture of nitrogen and helium gas is used to study internal natural convection up to 650?C. Similarly many other experimental studies have emphasized on HTGR core and plenum-to-plenum thermal-hydraulic behavior but with inert gases such as nitrogen or non-oxidizing solid substitutes for graphite. There is limited experimental data of mixed oxidation effects upon air ingress in nuclear grade graphite leading changes to thermal properties, heat generation due to chemical reactions and natural convection of air-helium mixture inside core at high temperatures (above 900?C). This work describes a unique experimental set-up developed by authors to conduct these mixed effect experiments. The current work is equally relevant to the designs well known by other names or geometries such as Very High Temperature Reactors (VHTRs) and Pebble Bed Modular Reactors (PBMRs). A common feature of HTGRs with passive safety design is that under certain accident conditions the decay heat is dissipated passively from the core by radiation/conduction/convection heat transport to the surrounding environment. Passive decay heat removal is one of the prime passive safety features of HTGRs and the heat conduction through the internal core region, graphite matrix, to the periphery is one of the major mechanisms of passive safety in HTGRs. The conduction and radiation heat transfer provide long-term heat removal but, during a Depressurized Loss of Forced Coolant (DLOFC) accident, the anticipated break in the coolant system will allow air ingress into the hot core region and can lead to graphite oxidation, which will negatively impact the core thermal behavior. In addition, air will also impact natural circulation of Helium and thus diminish the heat removal rate, which can further increase the graphite temperature. Therefore, both air ingress can eventually give rise to thermal instabilities and then to possible situations for breaching the safety envelope of TRISO fuel. This research summary is organized to present the background on oxidation studies, description of experimental setup and discussion of results in the following sections. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Study of Tritium Transport in Nuclear-Grade Graphite and Molten Fluoride Salt Systems

High-temperature salts are proposed as coolant in recent reactor technologies including Fluoride Salt-Cooled High Temperature Reactor (FHR) with fluoride salt as primary and secondary loop coolant, Molten Salt Reactors (MSRs) with fuel salt, and high-magnetic-field fusion machines with salt coolant blanket. The baseline coolant is {sup 7}Li{sub 2}BeF{sub 4} (flibe) due to its good thermal-hydraulic and neutronic properties. Tritium produced in flibe due to neutron irradiation makes the issue of tritium transport an area of great interest. In the temperature range of 427- 704 deg. C, tritium readily diffuses through metallic reactor structural materials, which makes tritium control and capture a key technical challenge. It is anticipated that tritium can be captured by graphite, as graphite material showed a high affinity for tritium in the Molten Salt Reactor Experiment (MSRE) due to the porous and irregular crystallite microstructure. Two types of nuclear grade graphite which used as core components for FHR: matrix graphite, which encapsulates the TRISO (Tri-structural-isotropic) fuel particles in FHR, and high purity, isotropic nuclear grade graphite IG-110 which can be used as moderator and reflector in reactor core for both fission and fusion reactor. The microstructure of matrix graphite and IG-110 is different from that of matrix graphite, due to the raw material and manufacturing processes. There are three phenomena important for predicting tritium transport in graphite-salt systems: tritium species diffusion in molten flibe; tritium species transport at the salt-graphite interface; and tritium species transport in graphite. In this summary microstructure characterization of IG-110 and matrix graphite, characterization of graphite-salt interface as well as techniques for trace tritium detection in graphite are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Post-Irradiation Examination for Conventional and Advanced Nuclear Fuel

During irradiation in a nuclear reactor, the fuel pellet experiences non-uniform fission rate and strong thermomechanical loads, which result in alterations and restructuring occurring non-homogeneously in the fuel volume. The irradiated fuel is characterized by a highly heterogeneous structure, with a correspondingly heterogeneous distribution of phases and properties. It is a big technical challenge for a hot cell facility to be able to cover effectively a broad variety of fuel types and compounds, characterized by different compositions, physico-chemical properties, geometries and configurations. The first challenge is to provide testing environments and procedures which do not alter or degrade the original irradiated fuel properties and are technically suitable for hot cell or shielded facility operation. For instance, non-oxide fuel compounds require protective atmosphere to avoid sample oxidation during storage, preparation and testing. Moreover, suitable adaptations of tools and measuring devices must be available to characterize fuel configurations which differ from the conventional cylindrical pellet-cladding geometries (e.g. spherical fuel elements for pebble bed HTR and TRISO coated fuel particles). Finally, adequate scientific methods of investigation must be available to measure properties and study the behaviour of highly heterogeneous irradiated fuel. In addition to basic techniques for non-destructive and destructive examination of nuclear fuel systems (e.g. covering fuel and cladding or fuel and coatings, or other configurations), 'in-depth' methods are applied to determine physical, thermomechanical and micro-analytical properties of specific phases/features present in irradiated fuel. This paper presents relevant trends and examples from fuel safety studies performed at JRC-Karlsruhe and in other laboratories which highlight the 'added value' obtainable by applying combined techniques, high resolution tools or 3D methods of analysis for the experimental 'multiscale' characterization of irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗