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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.

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All sourcesNASA NTRSDOE OSTI
NASA NTRS1989

Fabrication of light-weighted Si/SiC lidar mirrors

Small models of lightweight Si/SiC lidar mirrors have been fabricated via a scalable and cost-effective chemical vapor deposition (CVD) process. These mirrors are about 7.5 cm in diameter and consist of a Si-cladded SiC faceplate and lightweight structure made of SiC. The lightweight structure consists of an outer hexagonal cell with six triangular inner cells. The mirrors were polished to a figure better than 1/5 of a wave and a finish of better than 5A rms. The CVD route to fabrication of mirrors is fast and has the potential to yield several mirrors in a few weeks time from a signel reactor.

Goela, Jitendra S.↗
NASA NTRS2005

Poco Graphite Mirror Metrology Report

Recently a lightweight mirror technology was tested at Marshall Space Flight Center's Space Optic Manufacturing Technology Center (MSFC, SOMTC). The mirror is a Poco Graphite CVD Si clad SiC substrate. It was tested for cryogenic (cryo) survivability to 20deg Kelvin in SOMTC's X-ray Calibration and Cryogenic Test Facility. The surface figure of the mirror was measured before and after cry0 cycling. The test technique and results are discussed.

Kester, Thomas J.↗
DOE OSTI2017

Modelling Advanced Technology Fuels

Interest in new fuel technologies remains high after more than six years since the incident at Fukushima-Daiichi. Significant programs at national institutions and private companies work to understand new materials in an effort to improve the accident tolerance of nuclear fuel. Recently, the community has begun to use the title “Advanced Technology Fuels” to communicate that current fuels already have a degree of accident tolerance and to indicate that accident tolerance is not the only driving force in new fuel technology development. Any new fuel material faces challenges for widespread adoption. One significant challenge is the lack of experimental data. Experimental programs at Halden and elsewhere are working to provide the necessary experimental data, but much remains to be done. Given this lack of data, the use of modeling tools that can provide guidance during the design process is attractive. Support of advanced technology fuel modeling has been a significant point of emphasis in the BISON project over the past few years. We demonstrate the modeling capability, and the adaptability, of BISON using analyses of FeCrAl cladding, U3Si2 fuel, coated Zircaloy cladding, and SiC composite cladding. These examples point out strengths and weaknesses of the concepts, clarify where experimental data would be particularly useful, and indicate the utility of BISON for modeling current and future advanced technology fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Development of LWR Fuels with Enhanced Accident Tolerance

Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15 N and U 3 Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3 Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3 Si 2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3 Si 2 /68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2 , UN, and U 3 Si 2 the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2 . Cold spray application of either the amorphous steel or the Ti 2 AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Downselection of Cladding Materials for Zirconium Hydride Moderator

This report complete Milestone M3.1.1 — “Cladding downselection: Team will report to ARPA-E on at least two down-selected cladding options—one metallic and one ceramic.” In this milestone, we systematically assess FeCrAl alloys and SiC-based cladding options for a zirconium hydride (ZrHx) moderator in terms of their neutronics, radiation stability, hydrogen permeability, chemical compatibility, and fabricability. The analysis shows that FeCrAl alloys are excellent candidate cladding materials for a ZrHx moderator given their proven radiation stability, acceptable compatibility with ZrHx, and industrially established fabricability. The relative high hydrogen permeability is manageable by introducing a thin layer of an Al 2 O 3 hydrogen permeation barrier on the external surface. The SiC-cladded ZrHx moderator is promising because of its outstanding neutronics performance, excellent radiation stability, and extremely low intrinsic hydrogen permeability. However, potential interactions of SiC and Zr and the challenge in achieving a hermetic SiC-based cladding require further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Microstructural and nanomechanical studies of PVD Cr coatings on SiC for LWR fuel cladding applications

The microstructure and nanomechanical properties of pure Cr coatings deposited on SiC with various physical vapor deposition (PVD) methods have been investigated for the mitigation of hydrothermal corrosion of SiC-SiC f (SiC fiber-reinforced SiC matrix) composite fuel cladding in light water reactors (LWR). Cr coatings (4–7 μm thick) were deposited on unbiased SiC substrates using six variants of magnetron sputtering processes: (i) standard (S-DCMS), (ii) pulsed (P-DCMS), (iii) ion-assisted (I-DCMS), and (iv) pulsed ion-assisted direct current magnetron sputtering (PI-DCMS), (v) high-power impulse magnetron sputtering (HiPIMS), and (vi) bipolar HiPIMS (B-HiPIMS). Microstructural characterization and nanoindentation testing were used to evaluate effects of deposition technique on coating microstructure and nanomechanical properties. The application of positive reverse pulses to the sputter target induced coarsening of the columnar grain structure, while ion bombardment of substrates during deposition promoted densification of the coating microstructure. Both effects are exemplified in B-HiPIMS deposition, resulting in a high-density microstructure with compressive residual stress. Material pile-up around nano-indents and plastic work during nanoindentation were analyzed to understand deformation behavior of the coatings. Finally, the results suggest the B-HiPIMS process to be among the promising methods for the deposition of Cr coatings on SiC-SiC f for LWR cladding application.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Hydrothermal Corrosion of SiC Materials for Accident Tolerant Fuel Cladding with and Without Mitigation Coatings

As a candidate material for accident-tolerant fuel cladding for light water reactors (LWR), SiCf–SiC composite materials possess many attractive properties. However, prior work has shown that SiC is susceptible to aqueous dissolution in LWR coolant environments. To address this issue, candidate coatings have been developed to inhibit dissolution. For this study, CVD SiC samples were prepared with Cr, CrN, TiN, ZrN, NiCr, and Ni coatings. Uncoated SiC and SiCf–SiC samples were also prepared. The samples were exposed for 400 h in 288 °C water with 2 wppm DO in a constantly-refreshing autoclave to simulate BWR–NWC. Cr and Ni coated samples lost less mass than the uncoated SiC sample, indicating an improvement in performance. The CrN coating resisted oxidation, but some of the coating was lost due to poor adhesion. The TiN coated sample gained significant mass due to oxidation of the coating. ZrN and NiCr coatings showed significant corrosion attack. SiCf–SiC ceramic matrix composite materials dissolved much faster than the CVD SiC sample, demonstrating the need for mitigation coatings if CMCs are to be used in LWRs. This work demonstrates the promise of Cr, Ni and CrN coatings for corrosion mitigation in LWRs, and shows that NiCr and ZrN are not promising coating materials.

Raiman, Stephen S.↗
DOE OSTI2016

Characterization of SiC-SiC accident tolerant fuel cladding after stress application

In this work, SiC-SiC tubes with representative LWR (water cooled reactor) cladding dimensions were subjected to cycles of out-of-pile thermal stress followed by quenching. Samples were held at elevated temperatures and dropped into a water bath to replicate the high thermal stresses generated during reflooding. The samples were then characterized to measure changes in strength, permeability, and structure. Results were evaluated both for quenching into room temperature and boiling water. The results show that SiC-SiC composite tubes quenched from 1000 Celsius degrees in air into a boiling water bath exhibited no increase in He leak rate through the cylinder wall up to a pressure difference of 34.5 kPa compared to the as-fabricated state. SiC-SiC composite tubes quenched from 300-400 C. degrees into room temperature water showed an increase in He leak rate compared to the as-fabricated state. We attribute this to the more severe thermal gradient generated through the wall thickness due to higher heat flux characteristic of nucleate boiling compared to film boiling. A persistent vapor film forms around the samples quenched into boiling water, whereas no vapor film forms for the samples quenched into room temperature water. Another important result is that not loss in strength was noted compared to unquenched samples and that the stresses generated by the quenching were not enough to cause fiber fracture.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Thermo-mechanical analysis of SiC and FeCrAl cladding behavior under a loss-of- coolant accident

Alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys, in order to improve the accident tolerance of light water reactor (LWR) fuel. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys and silicon carbide composite materials that exhibit much slower oxidation kinetics in high-temperature st eam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely.

36 MATERIALS SCIENCE↗
DOE OSTI2013

SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points and clad configurations. The 2-ply sleeve samples show a higher bend momentum compared to those of the 1-ply sleeve samples. This is applicable to both the hybrid mock-up and bare SiC-CMC sleeve samples. Comparatively both the 1- and 2-ply hybrid mock-up samples showed a higher bend stiffness and strength compared with the standard Zr-4 mock-up sample. The characterization of the hybrid mock-up samples showed signs of distress and preliminary signs of fraying at the protective Zr-4 sleeve areas for the 1-ply SiC-CMC sleeve. In addition, the microstructure of the SiC matrix near the cracks at the region of highest compressive bending strain shows significant cracking and flaking. The 2-ply SiC-CMC sleeve samples showed a more bonded, cohesive SiC matrix structure. This cracking and fraying causes concern for increased fretting during the actual use of the design. Tomography was proven as a successful tool to identify open porosity during pre-test characterization. Although there is currently insufficient data to make conclusive statements regarding the overall merit of the hybrid cladding design, preliminary characterization of this novel design has been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high heat flux configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. These rabbits contain a wide variety of specimens including monolith tubes, SiC fiber SiC matrix (SiC/SiC) composites, duplex specimens (inner composite, outer monolith), and specimens with a variety of metallic or ceramic coatings on the outer surface. The rabbits are targeted for insertion during HFIR cycle 475, which is scheduled for September 2017.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Severe accident modeling of a PWR core with different cladding materials

The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermal Conductivity and Diffusivity for SiC Fibers for Use in ATF Cladding Composites

As part of the accident tolerant fuels (ATF) initiative, there is a need to understand the thermophysical properties of cladding for these fuels. One potential cladding material is silicon carbide (SiC) ceramic matrix composite reinforced with SiC fibers because of the corrosion resistance, resistance to hydrogen formation, high bulk thermal conductivity, and favorable mechanical properties of SiC. However, data about the thermal properties of individual SiC fibers in these composites is absent in the literature. This paper presents the measured axial thermal conductivity and thermal diffusivity of a leading SiC fiber (HiNicalon Type S) via a thermal measurement technique (transient electrothermal technique, TET) optimized for thin fibers. The results show that the measured properties are significantly lower than bulk values for SiC. Because of the low neutron absorption, as well as excellent thermal and mechanical properties, zircaloy has primarily been used for the fuel cladding in many nuclear reactors. A major drawback of the zircaloy is its hydrogen generation vulnerability. This issue has been known for some time, but the explosion of hydrogen accident in the Fukushima Daiichi nuclear disaster has motivated efforts for accident tolerant fuels (ATF) and their cladding as suitable substitutes for zircaloy. Fuel cladding made of silicon carbide (SiC) ceramic matrix composite (CMC) is an attractive alternative to the zircaloy. Investigation on the tubular monolithic a-SiC polytropes indicates that it has superior oxidation (hydrogen production) resistance (three orders of magnitude slower) than zircaloy. Additionally, it has well-documented, favorable mechanical properties, along with high temperature corrosion resistance. However, its bulk thermal properties can vary significantly depending on polytrope, the grain size, grain boundary, or impurities. At room temperature, SiC has a thermal conductivity as high as 490 Wm{sup -1}K{sup -1}, where the {sup 6}H polytrope has the highest conductivity. Thermal measurements of the SiC CMC are usually accomplished by laser flash method for thermal diffusivity (a) and thermal conductivity is indirectly calculated through the relationship k=aρcp. Depending on the manufacturing method and the reinforced fiber, the composite thermal conductivity varies from <10 Wm{sup -1}K{sup -1} to ∼70 Wm{sup -1}K{sup -1}. While the flash method is convenient for the measurement of the composite, it is difficult to apply to the measurement of individual fibers. Because of this, the thermal properties of individual SiC fibers is not present in the literature. Because the SiC fiber manufacturing process is different than that of the SiC matrix in the CMC, the properties of the bulk material might be erroneous when applied to the fiber. Being able to measure the fiber properties independently and directly allows for an understanding of the effect of the manufacturing process on the fiber and the opportunity to potentially improve or control the thermal properties of the composite. This improved accuracy can be achieved by measuring the fiber with a technique optimized for individual fiber measurement, known as the Transient Electrothermal Technique or TET. This paper presents the thermal conductivity and diffusivity of a leading SiC fiber (HiNicalon Type S). Measurements are performed with the TET, with the resulting properties being significantly lower than the bulk values of SiC, which is one possible explanation for relatively low CMC values. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of the effects of neutron irradiation on first-generation corrosion mitigation coatings on SiC for accident-tolerant fuel cladding

In this work, high purity SiC and SiC/SiC composites coated with commercial TiN, Cr, CrN, or CrN/Cr multilayer coatings were irradiated in Ar or flowing PWR water in the Massachusetts Institute of Technology Nuclear Reactor Laboratory (MITR). Irradiation in Ar was performed in the core. In the water environment, identical samples were placed in one of three different locations: in-core, providing exposure to neutron damage and radiolysis-affected water; above-core, where samples were exposed to radiolysis-affected water but not neutron damage, or outside of the core, where samples were exposed to the coolant water without the effects of radiation. Radiation in Ar revealed significant cracking of all but the TiN coatings, attributed to differential swelling between the coating and substrate. Lattice swelling was not observed in any of the coatings, but 0.2% void swelling was observed in the Cr coating. All of the coatings failed during water exposures in the core. Likewise, CrN/Cr spalled in each condition. Cr was protective, except under radiation damage as a result of cracking, and TiN severely degraded in the core with no coating was found following exposure. A SiC/coating ATF cladding system is anticipated to perform adequately following improvements in coating ductility and purity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Development of Accident Tolerant Fuel Options For Near Term Applications

In this section, the basis for the proposal followed by description of the work scope and methodologies are discussed. Following the Fukushima disaster in 2011, US congress mandated the department of energy to start a R&D program on accident tolerant fuel (ATF) concepts for the existing light water reactor fleet. By 2015, the start date of this IRP, several ATF concepts were being pursued by various entities. In general, the ATF concepts were divided into two categories: near term and long term. The near term concepts included: Coated Zircaloy clad, fuels with additives and dopants and FeCrAl/steel based claddings. The long term concepts included: SiC composite cladding, high density fuels (U 3 Si 2 , UN) and TRISO type fuel forms (e.g. FCM). It is of particular importance for any DOE funded effort to give background information and original Funding Opportunity Announcement (FOA) goals in its final report to evaluate whether the FOA goals and the proposed work were actually addressed. The FOA for this IRP sought for modeling and simulation capability to predict various ATF performances during normal operation, design basis accident and severe accident conditions. Particularly, the FOA was concern with the state of the core performance during a severe accident. It was postulated that if an ATF retains its structural integrity, the performance of non-fuel structures in the core needs to be addressed. The IRP team focused the proposal on Pressurized Water Reactor (PWR) performance for near term ATF concepts: Coated cladding, Fuel with additives and dopants to accelerate adoption of such concepts and address the unanswered questions regarding their performance. The project also gave limited focus on another near term concept, FeCrAl cladding, by restricting its scope to time-to-failure prediction since substantial effort supported by other programs including ORNL ATF program, GEH ATF program and NEAMS High Impact Problem were already made or underway during the IRP tenure. Despite two previous IRPs funded in 2012 with focus on experimental study of metallic and ceramic Zircaloy coated claddings, this IRP still included a coating program. This was a testament to the project leadership’s belief that the coated clad concept will be the primarily initial pathway for near term industrialization of ATF technology due to presence of enrichment penalty and licensing hurdles of the alternative clad concepts. In doing so, the IRP was positioned to provide key findings on coated Zircaloy as a uniquely funded external program to DOE’s industrial ATF campaign. Currently, in 2019, the coated clad concept is being pursued by all three vendors for licensed application and Lead Test Rods and Lead Test Assemblies (LTRs/LTAs) have already been inserted or planned to be inserted in the existing fleet. At the start of the IRP project, based on the meeting with the projects technical and federal point of contacts, it was also agreed upon if the contractual obligations are met, then the scope of the simulation work will be extended to Boiling Water Reactors as included in Section 2.10.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Radiation tolerant joining for silicon carbide-based accident tolerant fuel cladding

Silicon carbide (SiC) and silicon carbide fiber reinforced, silicon carbide matrix (SiC-SiC) composites are candidate materials for nuclear fuel cladding in accident tolerant light water reactors and advanced high temperature reactors due to their chemical stability and retention of strength at elevated temperature and neutron flux. A reliable means to join silicon carbide components is imperative to the successful implementation of SiC-based structural nuclear materials as fuel cladding as well as more complex core internal structures. A comprehensive approach has been taken to address several of the significant challenges associated with SiC joining and fabrication of complex SiC structures. This work focuses on low activation joining materials that are expected to behave well under a variety of different reactor operating conditions. Pre- and post-irradiation mechanical testing and microstructural evaluation will be presented for specimens subjected to fast flux irradiation to 4.5 dpa at 730 C. degrees, as well as 8.7 dpa at 270 C. degrees. Joints did not exhibit any strength degradation under either irradiation condition. Additionally, the performance of joined cylindrical SiC-SiC composite specimens was evaluated following irradiation in pressurized water reactor (PWR) coolant chemistry to 0.5 dpa. Lastly, out-of-pile characterization of permeability has been conducted on joined cylindrical SiC-SiC specimens to assess performance of this joint material under additional conditions relevant to reactor operation. Details of the experimental set-up and relationships between microstructure and performance will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗