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At least 847 records · Page 47
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DOE OSTI2020

Materials Data on CaBi2 by Materials Project

Bi2Ca1 is Zirconium Disilicide structured and crystallizes in the orthorhombic Cmcm space group. The structure is three-dimensional. Ca is bonded in a 10-coordinate geometry to ten Bi atoms. There are a spread of Ca–Bi bond distances ranging from 3.38–3.71 Å. There are two inequivalent Bi sites. In the first Bi site, Bi is bonded in a 4-coordinate geometry to six equivalent Ca and two equivalent Bi atoms. Both Bi–Bi bond lengths are 3.22 Å. In the second Bi site, Bi is bonded in a 8-coordinate geometry to four equivalent Ca and four equivalent Bi atoms. All Bi–Bi bond lengths are 3.34 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Hf3AlGe5 by Materials Project

Hf3AlGe5 is Zirconium Disilicide-derived structured and crystallizes in the orthorhombic Amm2 space group. The structure is three-dimensional. there are six inequivalent Hf sites. In the first Hf site, Hf is bonded in a 10-coordinate geometry to ten Ge atoms. There are a spread of Hf–Ge bond distances ranging from 2.78–3.20 Å. In the second Hf site, Hf is bonded in a 10-coordinate geometry to two equivalent Al and eight Ge atoms. Both Hf–Al bond lengths are 2.92 Å. There are a spread of Hf–Ge bond distances ranging from 2.75–3.12 Å. In the third Hf site, Hf is bonded in a 10-coordinate geometry to two equivalent Al and eight Ge atoms. Both Hf–Al bond lengths are 2.93 Å. There are a spread of Hf–Ge bond distances ranging from 2.74–3.06 Å. In the fourth Hf site, Hf is bonded in a 10-coordinate geometry to ten Ge atoms. There are a spread of Hf–Ge bond distances ranging from 2.78–3.20 Å. In the fifth Hf site, Hf is bonded in a 10-coordinate geometry to two equivalent Al and eight Ge atoms. Both Hf–Al bond lengths are 2.95 Å. There are a spread of Hf–Ge bond distances ranging from 2.74–3.05 Å. In the sixth Hf site, Hf is bonded in a 10-coordinate geometry to two equivalent Al and eight Ge atoms. Both Hf–Al bond lengths are 2.94 Å. There are a spread of Hf–Ge bond distances ranging from 2.75–3.11 Å. There are two inequivalent Al sites. In the first Al site, Al is bonded in a 8-coordinate geometry to four Hf and four equivalent Ge atoms. All Al–Ge bond lengths are 2.69 Å. In the second Al site, Al is bonded in a 8-coordinate geometry to four Hf and four equivalent Ge atoms. All Al–Ge bond lengths are 2.69 Å. There are ten inequivalent Ge sites. In the first Ge site, Ge is bonded in a 8-coordinate geometry to four Hf and four equivalent Ge atoms. All Ge–Ge bond lengths are 2.69 Å. In the second Ge site, Ge is bonded to four Hf and four equivalent Al atoms to form a mixture of distorted corner and face-sharing GeHf4Al4 hexagonal bipyramids. In the third Ge site, Ge is bonded to four Hf and four equivalent Al atoms to form a mixture of distorted corner and face-sharing GeHf4Al4 hexagonal bipyramids. In the fourth Ge site, Ge is bonded in a 8-coordinate geometry to four Hf and four equivalent Ge atoms. In the fifth Ge site, Ge is bonded in a 4-coordinate geometry to six Hf and two equivalent Ge atoms. Both Ge–Ge bond lengths are 2.67 Å. In the sixth Ge site, Ge is bonded in a 4-coordinate geometry to six Hf and two equivalent Ge atoms. Both Ge–Ge bond lengths are 2.67 Å. In the seventh Ge site, Ge is bonded in a 4-coordinate geometry to six Hf and two equivalent Ge atoms. Both Ge–Ge bond lengths are 2.70 Å. In the eighth Ge site, Ge is bonded in a 4-coordinate geometry to six Hf and two equivalent Ge atoms. In the ninth Ge site, Ge is bonded in a 4-coordinate geometry to six Hf and two equivalent Ge atoms. In the tenth Ge site, Ge is bonded in a 4-coordinate geometry to six Hf and two equivalent Ge atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on TiSi2 by Materials Project

TiSi2 is Zirconium Disilicide structured and crystallizes in the orthorhombic Cmcm space group. The structure is three-dimensional. Ti is bonded in a 10-coordinate geometry to ten Si atoms. There are a spread of Ti–Si bond distances ranging from 2.59–2.82 Å. There are two inequivalent Si sites. In the first Si site, Si is bonded in a 8-coordinate geometry to six equivalent Ti and two equivalent Si atoms. Both Si–Si bond lengths are 2.41 Å. In the second Si site, Si is bonded in a 8-coordinate geometry to four equivalent Ti and four equivalent Si atoms. All Si–Si bond lengths are 2.52 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on MgSi2 by Materials Project

MgSi2 is Zirconium Disilicide structured and crystallizes in the monoclinic P2_1/m space group. The structure is three-dimensional. there are two inequivalent Mg sites. In the first Mg site, Mg is bonded in a 12-coordinate geometry to two equivalent Mg and ten Si atoms. Both Mg–Mg bond lengths are 3.03 Å. There are a spread of Mg–Si bond distances ranging from 2.76–3.04 Å. In the second Mg site, Mg is bonded in a 12-coordinate geometry to two equivalent Mg and ten Si atoms. There are a spread of Mg–Si bond distances ranging from 2.77–3.04 Å. There are four inequivalent Si sites. In the first Si site, Si is bonded in a 10-coordinate geometry to six Mg and four Si atoms. There are two shorter (2.56 Å) and two longer (2.80 Å) Si–Si bond lengths. In the second Si site, Si is bonded in a 4-coordinate geometry to four Mg and two equivalent Si atoms. In the third Si site, Si is bonded in a 4-coordinate geometry to four Mg and two equivalent Si atoms. Both Si–Si bond lengths are 2.79 Å. In the fourth Si site, Si is bonded in a 10-coordinate geometry to six Mg and four Si atoms. Both Si–Si bond lengths are 2.56 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on HfScGe4 by Materials Project

ScHfGe4 is Zirconium Disilicide-derived structured and crystallizes in the orthorhombic Amm2 space group. The structure is three-dimensional. Sc is bonded in a 10-coordinate geometry to ten Ge atoms. There are a spread of Sc–Ge bond distances ranging from 2.79–3.12 Å. Hf is bonded in a 10-coordinate geometry to ten Ge atoms. There are a spread of Hf–Ge bond distances ranging from 2.79–3.13 Å. There are four inequivalent Ge sites. In the first Ge site, Ge is bonded in a 4-coordinate geometry to four equivalent Sc, two equivalent Hf, and two equivalent Ge atoms. Both Ge–Ge bond lengths are 2.61 Å. In the second Ge site, Ge is bonded in a 4-coordinate geometry to two equivalent Sc, four equivalent Hf, and two equivalent Ge atoms. In the third Ge site, Ge is bonded in a 8-coordinate geometry to two equivalent Sc, two equivalent Hf, and four equivalent Ge atoms. All Ge–Ge bond lengths are 2.71 Å. In the fourth Ge site, Ge is bonded in a 8-coordinate geometry to two equivalent Sc, two equivalent Hf, and four equivalent Ge atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on TmSn2 by Materials Project

TmSn2 is Zirconium Disilicide structured and crystallizes in the orthorhombic Cmcm space group. The structure is three-dimensional. Tm is bonded in a 10-coordinate geometry to ten Sn atoms. There are a spread of Tm–Sn bond distances ranging from 3.14–3.43 Å. There are two inequivalent Sn sites. In the first Sn site, Sn is bonded in a 4-coordinate geometry to six equivalent Tm and two equivalent Sn atoms. Both Sn–Sn bond lengths are 3.01 Å. In the second Sn site, Sn is bonded in a 8-coordinate geometry to four equivalent Tm and four equivalent Sn atoms. All Sn–Sn bond lengths are 3.09 Å.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Inferring nuclear structure from heavy isobar collisions using Trajectum

Nuclei with equal number of baryons but varying proton number (isobars) have many commonalities, but differ in both electric charge and nuclear structure. Relativistic collisions of such isobars provide unique opportunities to study the variation of the magnetic field, provided the nuclear structure is well understood. In this work we simulate collisions using several state-of-the-art parametrizations of the ^{96}_{40} 40 96 Zr and ^{96}_{44} 44 96 Ru isobars and show that a comparison with the exciting STAR measurement [arXiv:2109.00131] of ultrarelativistic collisions can uniquely identify the structure of both isobars. This not only provides an urgently needed understanding of the structure of the Zirconium and Ruthenium isobars, but also paves the way for more detailed studies of nuclear structure using relativistic heavy ion collisions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

SCALE Lattice Physics Code Assessments of Accident Tolerant Fuel

This report highlights accident-tolerant fuel (ATF) code assessment activities performed under Project NRCHQ- 60-17-T-0017, Lattice Physics Enhancements and Assessment. ATF covers a broad range of advanced fuel and clad designs for light water reactors (LWRs) to enhance performance under several accident conditions. Several ATF concepts are anticipated to be deployed as lead test rods (LTRs) or lead test assemblies (LTAs) within the next five years. The purpose of this work is to assess the predictive capabilities of NRC neutronics codes that underpin various licensing calculations. ATF designs use different fuel and clad materials compared to standard UO 2 fuel and zirconium-alloy claddings (hereafter UO 2 -Zry). These new materials and geometrical designs need to be assessed to quantify the impact of nuclear data uncertainties on quantities of interest (QOIs) in licensing calculations and the impact of modeling approximations which may be valid for UO 2 -Zry but not for ATF. This report outlines a systematic approach for ATF neutronics code assessment which includes sensitivity and uncertainty (S/U) analysis of nuclear data, identification of experimental benchmark and gaps for code validation, investigation of modeling approximations, and code-to-code comparisons of calculated QOIs against high-fidelity reference continuous energy (CE) Monte Carlo (MC) calculations. This report focuses on the assessment of the SCALE/Polaris lattice physics code for reactor safety analysis. Polaris lattice physics calculations generate few group (FG) cross sections for PARCS full-core calculations. (Full-core analysis, spent fuel analysis, and severe accident analysis will be performed in future work.) The selected ATF concepts for this report include Cr 2 O 3 and Al 2 O 3 -Cr 2 O 3 -doped UO 2 fuel, U 3 Si 2 fuel, FeCrAl cladding, SiC cladding, and Cr-coated cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask

This report provides a best-estimate evaluation of residual water content (post-dry out) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask, and evaluates the radiolysis of the residual free water, and the physisorbed and chemisorbed waters on the surfaces of the fuel and cask internal contents. The evaluation of radiolytic breakdown of those waters with gamma radiation causing the generation of hydrogen gas (H 2 ) is made using available literature data and models. This evaluation is part of the overall materials performance evaluation of the SNF-in-canister system, and is part of the technical bases for their continued safe dry storage. The TN-32 cask contents included 32 HBU LWR spent fuel assemblies each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless steel structural components. The residual free and surface (physisorbed/chemisorbed) waters are ascribed to water vapor in the free volume and to components’ surfaces, respectively. The total potential radiolytic hydrogen inventory from the water vapor and from waters ascribed to surfaces has been calculated assuming all the water produced molecular H 2 . The residual water that is chemically incorporated into the bulk of a hydrated oxide, i.e., chemisorbed water, and its total potential hydrogen inventory has been calculated. These calculations are at the physical limit of material available and are used for a bounding assessment purpose only.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

MPA Materials Matter Summer 2020 [Newsletter]

The summer issue of the Materials Physics and Applications division (MPA) newsletter features the following short articles: recognition for Rod Borup, From Wolfgang's Desk (Wolfgang Runde), From Adam's Desk (Adam Rondinone), ''Frustrating harmful dendite formation in potassium ion batteries''; ''From a Mott insulator to an ideal two-dimensional metal''; ''First ever actinide-zirconium framework synthesized''; and finishing with the usual service celebrations and HeadsUp sections.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Best Practices for Shielding Analyses of Activated Metals and Spent Resins from Reactor Operation

This report investigated best practices for performing shielding evaluations of Type B waste packages, as defined in 10 CFR Part 71 regulations on packaging and transportation of radioactive material, or packages for which the contents are not defined before loading and may include a broad range of nuclides, geometries, and non-fuel materials. The following non-fuel waste streams were analyzed: Activated metals from decommissioned commercial power reactors, including Type 304 stainless steel, reactor vessel steel, and Inconel, Control blades from boiling water reactors, Neutron-activated corrosion products on surfaces of activated metals, and Spent resins from power plant operations. Measured elemental compositions, including major constituents and impurities, for steel and Inconel samples from commercial power reactors were used in activation calculations to determine radionuclide inventories in activated metals. For a simplified cask model, 60 Co contribution to the total external package dose rate at 30 days after shutdown varied from approximately 60% to 95%, depending on the activated metal, initial cobalt impurity concentration in the metal, and the thickness of the overpack gamma shield. Its maximum contribution to the total external package dose rate of approximately 100% was reached within the time interval of 2 to 5 years after shutdown and was maintained for up to 45 to 60 years after shutdown, depending on material, initial cobalt impurity concentration, and shield thickness. Thereafter, the 60 Co contribution to external package dose rate decreased with increasing decay time. Cobalt-60 is primarily produced by neutron reactions with the cobalt impurity in steel and Inconel. Other important radionuclides in activated metals contributing to package external dose rate are radionuclides with relatively short decay times, including 51 Cr, 59 Fe, 58 Co, and 54 Mn. These radionuclides may be represented as an equivalent 60 Co activity/source because 60 Co gamma ray emissions are bounding in terms of source strength and energy to other important radionuclides identified in the analyzed activated metals. Approaches for modeling the neutron-activated corrosion products that may be attached to activated reactor components were analyzed in this report. It was demonstrated that a surface source is more conservative than a uniform volumetric source for the treatment of neutron-activated corrosion products with respect to external package dose rates. An analysis of the maximum radionuclide loadings reported on spent resins identified the radionuclides 137 Cs, 60 Co, 134 Cs, 65 Zn, and 58 Co as the primary contributors to external package dose rate. For a resin cooled for 3.08 years, the external package dose rate was entirely produced by the reported 137 Cs and 60 Co inventory. The neutron sources from actinides found on spent resins or activated metals produced negligible dose rates and may be ignored in dose rate analyses. Effects of idealized waste material, source geometry, and spatial material/source distributions on external package dose rates were determined based on dose rate results for a simplified cask model under normal conditions of transport. Type 304 stainless steel, zirconium, and aluminum with adjusted mass densities based on a maximum content weight were analyzed for material modeling effects on external package dose rate. These materials produced identical external package dose rates within the statistical uncertainties of the dose rate estimates. Among four different source geometry configurations with homogeneous material of different mass densities, uniform volumetric source distribution, and the same source strength, the geometry configuration with lowest mass density (i.e., minimum self-shielding effect among the four cases) was most conservative. Spatial source distributions that better represent localized peak 60 Co activity values were more conservative than a uniform volumetric source distribution, assuming the same weight and total source strength per package. The increase in external dose rate caused by localized source peaks can be as much as the ratio between source peak activity density to average activity density, depending on the location of the activated metal with peak activity density. Therefore, the shielding analysis may be simplified if localized peak activities and the average activity per package can be measured/determined and documented at the time of cask loading. For simplicity, external package dose rates may be determined based on average source activity and a uniform volumetric source distribution. The dose rate results from that calculation model multiplied by the ratio of peak activity density to average activity density will produce maximum dose rate values for conservative estimates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

IER-519 CED-2: Final Design for Thermal/Epithermal eXperiments (TEX) with Absorbers to Provide Validation Benchmarks for Hanford Tank Farms

The Hanford tank farms contain 56 million gallons of waste across 177 tanks. The primary criticality safety concern for the waste is the plutonium inventory in waste solids – approximately 670 kg in total. Criticality safety analysis credits the absorption and dilution properties of the large quantities of other elements (aluminum, chromium, iron, manganese, nickel, silicon, sodium, and zirconium) present in the waste. Of these, iron and manganese are by far the most significant neutron absorbers, particularly for the waste compositions of highest criticality safety concern. The criticality safety analyses at the Hanford Waste Treatment and Vitrification Plant (WTP) and the Savannah River tank farms also credit iron and manganese as the primary neutron absorbers to demonstrate subcriticality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Test Plan for Investigation of Hydrogen Generation in SNF Canisters

This report presents a test plan to investigate radiolytic hydrogen generation in spent nuclear fuel (SNF) canisters containing Zr-based cladding materials. The initial primary contributor to the generation of radiolytic hydrogen is estimated to be the residual physisorbed water on the ZrO 2 film, post-dry out. Over the longer term the primary hydrogen source is associated with the aluminum hydroxides and water vapor. The proposed testing to be conducted in FY22 would consist of lab-scale testing on non-radioactive surrogates of oxidized ZIRLO tubing and unalloyed zirconium using a miniature steel canister (“mini-canister") designed to allow intermittent, in-situ sampling of the canister gas during irradiation. The mini-canister approach for irradiation with in-situ monitoring system was used in a recent and on-going study [McNamara and Verst, 2020a & 2020b; Verst et al., 2021] to measure radiolytic hydrogen generation from aluminum SNF cladding surrogate materials. This proposed testing will form part of the overall materials performance evaluation of the commercial SNF-in-canister system and is part of the technical bases for their continued safe dry storage

08 HYDROGEN↗
DOE OSTI2021

High Level Gap Analysis for Accident Tolerant and Advanced Fuels for Storage and Transportation

This initial gap analysis considers proposed accident tolerant fuel (ATF) options currently being irradiated in commercial reactors, since these are most likely for future batch implementation. Also, advanced fuel (AF) options that may be likely for use in advanced reactors are considered. The cladding technologies considered were chromium-coated zirconium-based alloys, FeCrAl, and both monolithic and matrix composite Silicide carbide (SiC). The fuel technologies considered were chromium-doped uranium dioxide fuel, uranium alloys, uranium nitride, and uranium silicide. Numerous national labs, industry, and countries are performing significant testing and modeling on these proposed technologies to establish performance, but at this time none of the prototypes being irradiated have achieved end-of-life (EOL) burnup. There are some testing results after one burnup cycle to verify in-reactor performance, but little data beyond that. As the ATF prototypes acquire more burnup, data will be produced that is relevant to storage and transportation. The DOE:NE Spent Fuel and Waste Science and Technology (SWFST) Storage and Transportation (ST) Control Account will evaluate the performance data as it becomes available for application to the identified gaps for ST.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fundamental Studies of Tritium Diffusivity in Pure and Defective Zircaloy-4 Getters (Tritium Science Program FY2020 Report)

Zirconium (Zr) alloys are extensively used as tritium (T) getters in nuclear reactors due to their low absorption cross section to thermal neutron, good mechanical properties at high temperature and resistant to corrosion in different environmental conditions. Zr alloys have better thermal properties than many other refractory alloys including stainless steel. The nuclear characteristic is simulated, and T is produced in tritium-producing burnable absorber rods (TPBAR) by the irradiation of neutron flux in pressurized water reactor (PWR). T thus produced diffuses through the pellets and is captured by the getter where it chemically reacts with Zr to form metal hydride (ZrTx). These hydrides are brittle and affect adversely the mechanical properties of alloys. In addition to that, mismatch in the lattice structures after hydride formation creates a stress which needs to be considered in component design and their life evaluation. Therefore, understanding the behavior of T and its species becomes significant as fuel burnup is increased, which leads to increase in hydrogen pickup and oxide formation. The transport of T and its species in pure Zr alloys, and in alloys with certain alloying elements and defects are important to understand in order to enhance the performance of the materials.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Synthesis and Exploratory Catalysis of 3d Metals: Atom and Group-Transfer Reactions and the Activation and Functionalization of Small Molecules Including Greenhouse Gases

Determining ways to convert natural gas, with zero emissions, and to more value-added materials such as olefins is one of the main goals in my research group. Over this funding period, we explored the chemistry of early-transition metals with metal-nitrogen multiple bonds, specifically titanium and zirconium nitrides, and explored their redox properties, basicity and reactivity with small molecules including greenhouses gases such as carbon dioxide. Some of these work serve as inspiration for the chemistry of important materials such as uranium nitride along with its unprecedented basicity and ability to activate C-H bonds. Using robust chelating templates we also examined rare examples of trivalent group 4 transition metal ions, and studied these spectroscopically. Using early transition metal ions we also explored synthetic routes to new phosphide based products using the phosphaethynolate salt. In addition, we also explored the chemistry of ferrous systems, and redox active ligand that can allow us to reversibly break and make C-S as well as N-N bonds, activate dinitrogen forming unusual MNNM topologies. Our final component describes how we activate methane and dehydrocouple if with a carbene source to form ethylene. Using this information, we also discovered a simple to make iridium catalyst that activate and functionalize methane with 9:1 selectivity for mono-functionalization. We have established a pathway that leads to poisoning of the catalyst and have found optimal conditions for higher selectivity and with over 170 Turnovers.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Hydrogen Diffusivity Measurements of YH 1.87 Moderator Material with Incoherent Quasielastic Neutron Scattering

Yttrium hydride is an optimal choice of moderator material for thermal neutron spectrum reactors requiring small core volumes. The Transformation Challenge Reactor (TCR) program at Oak Ridge National Laboratory (ORNL) aims to develop an additively manufactured microreactor core by leveraging recent advances in materials, manufacturing, data analytics, and high-fidelity modeling and simulation. Yttrium hydride was selected as the moderator material for TCR due to its high moderating power, which enables the spatially efficient reactor design and thermal stability that is superior when compared with zirconium hydride. Hydrogen desorption from yttrium hydride is expected at elevated temperatures, posing safety and performance concerns. As a lifetime component in advanced reactors, appropriate strategies to mitigate hydrogen release from yttrium hydride over long-term reactor operation are required. Understanding hydrogen mobility in yttrium hydride provides the scientific basis for developing a hydrogen barrier for hydride moderator and predicting the hydrogen redistribution within the material during various operational conditions. This study investigated the hydrogen motion in YH 1.87 as a function of temperature ranging from 200 to 1,173 K using incoherent quasielastic neutron scattering (IQNS) at the Spallation Neutron Source (SNS). The results are presented and discussed in terms of hydrogen self-diffusion coefficients, activation energy for translational hydrogen motion, and hydrogen jump distances. Based on the IQNS data, YH 1.87 is found to be stable at high temperatures with no translational H motion below 1,023 K within the IQNS detection limits. In the temperature range of 1,073–1,173 K, hydrogen diffusivity in YH 1.87 is $D = 4.57 X 10^{⁻3} [m^2/s] exp (⁻\frac{1.73[eV]}{RT}$). The hydrogen concentration’s dependence on hydrogen diffusivity is also discussed in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Report on Thermal Cycling Testing and Bimaterial Weld Development for a Passively Actuated Materials Surveillance Test Article

This report describes the continuation and conclusion of a thermal cycling test program designed to evaluate a family of passively actuated material surveillance test articles. Additionally, the report details a preliminary study of bimetallic welding between 316 stainless steel and TZM (tungsten-zirconium-molybdenum) alloy. The test program was ultimately successful in demonstrating the basic viability of the passively actuated test article concept as well as specifically validating the procedures developed to size specimens to mimic the mechanical response of operating components. The report discusses and summarizes the results of the complete test program which tested samples for more than 1,000 controlled thermal cycles. The weld development program included both stir friction and electron beam joining. The stir friction welding was ultimately successful, while the electron beam welding would require additional development. Ultimately, these test programs pave the way for implementing a materials surveillance program in future molten salt reactors (MSRs). The report conclusions outline a plan for transforming the research and developed described here into a definite set of practices and procedures for a surveillance program at a future operating reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗