Advances in Ceramics for Environmental, Functional, Structural, and Energy Applications

By J.P. Singh

This volume contains 20 manuscripts presented during the Materials Science & Technology 2017 Conference (MS&T’17), held October 8-12, 2017 at the David L. Lawrence Convention Center, Pittsburgh, PA. Papers from the following symposia are included in this volume:

• 9th International Symposium on Green and Sustainable Technologies for Materials Manufacturing and Processing
• Advances in Dielectric Materials and Electronic Devices
• Construction and Building Materials for a Better Environment
• Innovative Processing and Synthesis of Ceramics, Glasses and Composites
• Materials Issues in Nuclear Waste Management in the 21st Century
• Materials Development for Nuclear Applications and Extreme Environments
• Materials for Nuclear Energy Applications
• Nanotechnology for Energy, Healthcare and Industry
• Processing and Performance of Materials Using Microwaves, Electric and Magnetic Fields, Ultrasound, Lasers, and Mechanical Work – Rustum Roy Symposium

These symposia provided a forum for scientists, engineers, and technologists to discuss and exchange state-of-the-art ideas, information, and technology on advanced methods and approaches for processing, synthesis, characterization, and applications of ceramics, glasses, and composites. 

Each manuscript was peer-reviewed using The American Ceramic Society’s review process. The editors wish to extend their gratitude and appreciation to their symposium co-organizers, to all of the authors for their valuable submissions, to all the participants and session chairs for their time and effort, and to all the reviewers for their comments and suggestions. 

We hope that this volume will serve as a useful reference for the professionals working in the field of materials science.

• Language: English
• Publisher: Wiley
• Release date: Oct 4, 2018
• ISBN: 9781119543282

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Preface

This volume contains 20 manuscripts presented during the Materials Science & Technology 2017 Conference (MS&T’17), held October 8-12, 2017 at the David L. Lawrence Convention Center, Pittsburgh, PA. Papers from the following symposia are included in this volume:

9th International Symposium on Green and Sustainable Technologies for Materials Manufacturing and Processing

Advances in Dielectric Materials and Electronic Devices

Construction and Building Materials for a Better Environment

Innovative Processing and Synthesis of Ceramics, Glasses and Composites

Materials Issues in Nuclear Waste Management in the 21st Century

Materials Development for Nuclear Applications and Extreme Environments

Materials for Nuclear Energy Applications

Nanotechnology for Energy, Healthcare and Industry

Processing and Performance of Materials Using Microwaves, Electric and Magnetic Fields, Ultrasound, Lasers, and Mechanical Work – Rustum Roy Symposium

These symposia provided a forum for scientists, engineers, and technologists to discuss and exchange state-of-the-art ideas, information, and technology on a dvanced methods and approaches for processing, synthesis, characterization, and applications of ceramics, glasses, and composites.

Each manuscript was peer-reviewed using The American Ceramic Society’s review process. The editors wish to extend their gratitude and appreciation to their symposium co-organizers, to all of the authors for their valuable submissions, to all the participants and session chairs for their time and effort, and to all the reviewers for their comments and suggestions.

We hope that this volume will serve as a useful reference for the professionals working in the field of materials science.

Morsi M. Mahmoud

Kumar Sridharan

Henry Colorado

Amar S. Bhalla

J. P. Singh

Surojit Gupta

Jason Langhorn

Andrei Jitianu

Navin Jose Manjooran

MATERIALS FOR NUCLEAR ENERGY APPLICATIONS

WESTINGHOUSE ACCIDENT TOLERANT FUEL MATERIALS

Frank Boylan¹, Peng Xu², Javier Romero², and Ed Lahoda³

¹Westinghouse Electric Company, Cranberry Township 16066

[phone]; [e-mail]

²Westinghouse Electric Company, Columbia, SC 29061

³Westinghouse Electric Company, Cranberry Township, PA 16066

ABSTRACT

Westinghouse is commercializing two unique accident tolerant fuels (ATFs): silicon carbide (SiC) as produced by General Atomics with uranium silicide (U3Si2) fuel and Cr coated zirconium alloy cladding with U3Si2 fuel. Testing of the cladding alternatives in autoclaves has been performed and samples have begun irradiation at the Massachusetts Institute of Technology Reactor and the Halden Project Reactor. Uranium silicide fuel is undergoing exposure in the Advanced Test Reactor and fuel pins have been removed and are undergoing post irradiation examination (PIE) at the Idaho National Laboratory (INL). This paper provides an update on these activities and a summary of results.

INTRODUCTION AND BACKGROUND

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program utilizes Cr coated zirconium alloy (CZA) cladding with U3Si2 high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. The lead test assembly (LTA) program will use both SiC/SiC composites from General Atomics and Cr coated zirconium alloy claddings with the high density/high thermal conductivity U3Si2 pellet which will begin in 2022. Over the past several years, Westinghouse has tested the Cr coated zirconium and SiC claddings in autoclaves and in the Massachusetts Institute of Technology (MIT) reactor and U3Si2 pellets in the Advanced Test Reactor (ATR). High temperature tests at the state-of-the-art facilities in Churchill, PA have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings.

WESTINGHOUSE ATF ACTIVITIES

Autoclave Corrosion Testing

Westinghouse has performed corrosion testing using the autoclave facility at the Churchill, PA site to screen various coatings and SiC preparation methods for corrosion resistance. As part of a multi-year program, over 12 types of coatings on zirconium alloys and approximately 10 versions of SiC have been tested in autoclaves. As a result of this testing, two coatings (Table I) were identified for testing in the MIT reactor.

Testing in the MIT reactor further narrowed the options to the Cr coating. Based on the positive test results, Westinghouse is now exploring methods for production of full length rods for LTRs to be constructed in 2018 for inclusion in a commercial reactor in early 2019.

Table I – Top Zirconium Alloy Coatings Autoclave Corrosion Performance At 360°C Water

Initial autoclave and reactor testing indicated relatively high levels of SiC corrosion. Autoclave testing with hydrogen peroxide was used to simulate more aggressive oxidation conditions of the reactor and to explore coolant conditions that would minimize SiC corrosion rates. The full battery of testing has been used to refine the manufacturing parameters of the SiC composites such that along with hydrogen addition to the primary coolant above 40 cc/kg [1], the current corrosion rates for SiC meet or exceed the target 7 microns/year recession rate. For a full core of SiC cladding, this would result in a maximum of 150 kg of SiO2 or about 300 ppm over an 18 month cycle. This is well below the solubility limit of ~700 ppm SiO2 at the coldest steam generator conditions. Note also, that resins are commercially available that could be added to the current resins used to maintain water chemistry to remove SiO2 on a continuous basis.

High Temperature Testing

The goal of the ATF program is to develop fuels that can withstand post-accident temperatures greater than 1200°C without the cladding igniting in steam or air. Therefore a crucial part of the testing carried out by Westinghouse over the previous year was aimed at quantifying the maximum temperature at which the ATF claddings could operate without excessive corrosion. The test apparatus first used current applied directly to the coated zirconium tubes. However, it was found that as the temperatures increased, issues with the connection of the test piece to the current source caused excessive resistance resulting in excessive heating and then burnout of the samples at the connection point. This direct heating method was then replaced with a graphite rod which was inserted into insulation and then into the test piece. This resulted in very stable heating of the test pieces.

CZAs have now been run at up to 1400°C. This is above the Cr-Zr low melting eutectic point of 1333°C. At 1400°C, there was noticeable reaction between the Cr and the Zr. However, there was not the rapid oxidation that uncoated Zr experiences at 1200°C, so that there is likely some reasonable residence time that the cladding could survive at temperatures above 1400°C.

At temperatures of 1300°C, the Cr coated zirconium alloy was stable for reasonable lengths of time. Combined with the lowering of zirconium oxidation at normal operating temperatures which vastly reduces the formation of zirconium hydrides and therefore embrittlement, the Cr coated zirconium has shown that it will provide significant improvement in the performance during normal operation, transients, design basis accidents and beyond design basis accidents as compared to uncoated zirconium.

Similar tests were run with SiC at temperatures from 1600°C up to 1700°C. These tests were run with the graphite heater rod and were terminated only because of excessive corrosion of the heater rod. At 1600°C, the SiC was visually untouched. At 1700°C, there were indications of small beads on the surface (presumably SiO2 from the reaction of SiC with steam) but on the whole, no significant deterioration of the SiC. Changes are being made to the heating rod to increase the flow of He cover gas and to allow accurate weight changes to be made on the SiC rodlets so that kinetic data can be obtained.

U3Si2 Testing

U3Si2 has never been utilized as pellets inside cladding in Light Water Reactor (LWR) fuel service. There is a lack of data on the behavior of U3Si2 at LWR operating temperatures (estimated to be from 600°C and up to 1200°C during transients). To remedy this lack of data, U3Si2 fuel pellets were manufactured at Idaho National Laboratory (INL) and put into rodlets in the ATR in 2015. The first rodlets came out of ATR at the end of 2016 and are due for destructive post irradiation examination in the summer of 2017 at INL [2]. Preliminary nondestructive testing from neutron radiography of the U3Si2 Pins after exposure of 20 MWd/kgU in the ATR shows very good results with a lack of pellet cracking and distortion.

U3Si2 was tested for air and steam oxidation as compared with UO2 using digital scanning calorimeters at both the Westinghouse Columbia facility [3] and at Los Alamos National Laboratory (LANL) [4]. The Westinghouse results indicate that the ignition temperatures for UO2 and U3Si2 are between 400°and 450°C (Table II). The LANL results (Table II) indicates an ignition temperature of about 400°C. The reasons for this difference are being studied. The reactivity of U3Si2 and UO2 are comparable at normal operating conditions (320°C), though the heat generated and mass generated by the oxidation of the U3Si2 is considerably higher than for UO2 at higher temperatures. The effect of this difference in heat release and mass on the stability of the rods is being investigated in rodlet tests in the Churchill autoclaves in the summer of 2017. However, the risk of any reactions between the coolant and the U3Si2 is probably much less than the current 1 to 2 ppm due to current rod failures, because the ATF claddings tend to be much harder than zirconium alloys. Therefore, it is expected that grid to rod fretting leakages will be eliminated.

Finally, LANL identified the potential for the formation of a U3Si2-H1.8 compound in the event of a leaker. Further work reported by S. Mašková et al [5] indicated that this would not likely be an issue since the operating temperature of the U3Si2 fuel will be above the decomposition temperature (~550°C) of this compound.

Table II – Comparison of Westinghouse U3Si2 and UO2 Scanning Calorimeter Results (Heating Rate 2.5°C/min)

CONCLUSION

The Westinghouse ATF concepts appear to be technically achievable as LTRs and LTAs in the 2019 and 2022 timeframe. Performance issues with SiC, coated cladding, and U3Si2 fuel have been identified and overcome through modifications, engineering, and testing. As with any revolutionary new product, technical challenges may surface, but the robust research and development program that Westinghouse has in place will be used to overcome these challenges.

REFERENCES

Ed Lahoda, Sumit Ray, Frank Boylan, Peng Xu and Richard Jacko, SiC Cladding Corrosion and Mitigation, TopFuel 2016, Boise, Idaho, Paper Number 17450, September 10, 2016.

Jason Harp, Idaho National Laboratory, private communication, preliminary examination.

Lu Cai, Peng Xu, Andrew Atwood, Frank Boylan, Edward J. Lahoda, Thermal Analysis of ATF Fuel Materials at Westinghouse, ICACC 2017, Daytona Beach, Florida, January 26, 2017.

E. Sooby Wood, J.T. White, A.T. Nelson, Oxidation behaviour of U-Si compounds in air from 25 to 1000 C, Journal of Nuclear Materials, 484 (2017) pages 245-257.

S. Mašková, K. Miliyanchuk, L. Havela, Hydrogen absorption in U3Si2 and its impact on electronic properties, Journal of Nuclear Materials, 487 (2017) pages 418-423.

ACKNOWLEDGEMENT

This material is based upon work supported by the Department of Energy under Award Number DE-NE0008222.

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

INVESTIGATIONS OF CAPACITOR DISCHARGE WELDING FOR THE ATTACHMENT OF ENDCAPS TO MOLYBDENUM-BASED NUCLEAR FUEL ROD CLADDING

Jerry E. Gould¹, Cem Topbasi², and Bo Cheng²

¹EWI

Columbus, OH, U.S.A.

²EPRI

Palo Alto, CA, U.S.A.

ABSTRACT

In response to the Fukushima nuclear power plant disaster there has been strong interest in fuel rod claddings with enhanced survivability during cooling loss. A strong candidate for such cladding is molybdenum. The high thermal stability of the material is key to improving cladding performance. A challenge to the use of such claddings is providing end sealing once the fuel is loaded. Molybdenum is susceptible to excessive grain growth during welding, resulting in a loss of ductility and toughness. Such grain growth can be minimized (or eliminated) by reducing both bonding temperature, as well as associated exposure times. Capacitor discharge (CD) welding offers significant potential for both. First, it is a solid-state process, functioning below the melting temperature of the substrates and, in addition, has implicit heating times in the range of milliseconds. In this study, CD welding was employed for attaching a range of endcap materials to molybdenum tubes. Endcap materials included Zircaloy-4, Kanthal, and molybdenum. Joint preparation was for an edge type weld, employing a taper on the endcaps and the tube wall as the projection. In these trials, relative weldability improved as the end caps were changed from Zircaloy-4 to Kanthal to molybdenum. Poorest combinations (Zircaloy-4 caps) showed excessive deformation of the endcap, with none on the edge of the tube wall. Best combinations (molybdenum endcaps) showed forging of the tube wall end into the opposing substrate. Best-identified practices included welding with molybdenum endcaps, use of a rapid (2.5 ms) current pulse, and a low relative welding force (2.2 kN). Resulting joints showed a solid-state character, no apparent grain growth, and leak tightness.

INTRODUCTION

On March 11, 2011, the nuclear power plant in Fukushima was disabled by a major tsunami¹. During the flooding associated with that tsunami, the facility lost all power resulting in a failure of the cooling systems for the reactor core. The remaining water in the core boiled and the resulting steam reacted with the zirconium-based cladding with additional heat generation, eventually resulting in melting of the core itself¹

Following this event, there has been a concerted industry effort to develop fuel cladding with enhanced survivability compared to the existing Zircaloy-4-based technology. Much of the focus has been on developing refractory metal-based cladding². The material of choice has been molybdenum. Molybdenum is a refractory metal, with a melting point of roughly 2600°C (compared to 1700°C for Zr) that can maintain its mechanical integrity to over 1500°C. Molybdenum, however, has several issues that must be addressed for use in this application. First, molybdenum has a neutron cross section considerably larger than that of Zirconium and comparable to that of steel³. This size difference affects the efficiency of the reactor, as well as can result in swelling of the clad itself. Also, molybdenum has corrosion concerns in the cooling water-based environment and is susceptible to oxidation at higher temperatures²,⁴. To address these concerns, a new generation, multi-layer cladding approach has been developed⁴. This cladding includes an interior layer of molybdenum, a thin transition layer of niobium, and an external layer of zirconium. The resulting product offers the opportunity for enhanced thermal stability (implying survivability), as well as environmental compatibility in the operating environment.

Fabrication of fuel rods, of course, requires attachment of endcaps. For this new generation of claddings, it is implied that the endcap itself is capable of being joined to the high-temperature stability (molybdenum) interior layer. The creation of joints, including a molybdenum element, is generally considered problematic. Refractory metals (including molybdenum) have a body-centered cubic crystal structure that remains unchanged as the material is heated to the melting point. As a result, during welding these materials can experience extensive grain growth. This grain growth has been related to increases in the ductile-to-brittle transition temperature, with associated loss of ductility⁵,⁶. Molybdenum is also seen to be incompatible with other candidate materials for fuel rod construction⁷. For example, with respect to zirconium, a eutectic can be formed resulting in a 300°C melting point suppression. This can result in liquation-related cracking of the resulting joint. In addition, on cooling, this combination can also result in Laves phase formation embrittling the joint. Kanthal is also a candidate material for fuel rods. This is an aluminum containing stainless steel designed for thermal stability. However, in attachment to molybdenum many intermetallic phases can form, potentially embrittling the joint.

Previous work has shown that solid-state processes can be an effective approach for the joining of molybdenum alloys⁸. Solid-state joining methods include technologies such as friction, flash-butt, and upset-butt welding. These processes offer some unique advantages compared to fusion technologies for joining refractory metals. These include lower (below melting) processing temperatures, deformation during joining, and relatively short thermal cycles.

CD welding fits into this class of solid-state processes⁹. This is a resistance-based process, using current flow through the workpieces to generate heat. That current is provided by one or more capacitors charged to relatively high (up to 3000 V) voltages. The current discharge itself is designed to be very rapid. Typical CD welding times can range from a few up to about 100 milliseconds. CD welding is typically done in specially designed load frames. Essentially, the workpieces are placed under an applied force and the current discharged through the stack-up. The rapid heating results in thermal softening of the substrates, with subsequent forging to create a joint. It is of note that the process is mechanically, as well as electrically, dynamic (the workpieces are forged together)⁹. Therefore, the system must be designed to match electrical and mechanical response for optimum joining.

In this work, CD welding has been applied for attaching endcaps to specimens representative of candidate molybdenum fuel rods. Work was conducted with three types of endcaps. These included those made from Zircaloy-4, Kanthal, and molybdenum. Work has included design of the joint geometry itself, tooling development, welding trials, and quality assessments.

EXPERIMENTAL PROCEDURES

As suggested above, endcaps for study were made from Zircaloy-4, Kanthal, and molybdenum. The tubes themselves were from molybdenum. Nominal compositions for each material, based on standard references, are provided in Table I¹⁰-¹¹. Tubes for joining included a nominal diameter of 9.5 mm, and a wall thickness of 210 μm. Tubes were cut to a nominal length of 22 mm for use as samples. Endcaps were all nominally 9.3 mm in diameter and 3.8-mm thick. All endcaps included a 45-degree taper over the last 1.3 mm of length on one end. This was done to facilitate an edge projection-type joint configuration. A sample endcap with taper is shown in Figure 1. The actual joint geometry prior to welding is provided in Figures 2 and 3. This joint geometry is referred to as an edge projection weld¹². It is commonly used in resistance welding for tube sealing