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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Performance Testing of Enhanced Linear Variable Integrated Sensor (ELVIS III) for LVDTs

The Enhanced Linear Variable Intrinsic Sensor (ELVIS) was a 2024 breakthrough discovery by Idaho National Laboratory (INL) in terms of further innovating the usage of all linear variable differential transformers (LVDTs) and enhancing their performance in harsh environments. In irradiation tests conducted within material test reactors (MTRs), LVDTs with an internal temperature sensing capability can address several critical challenges. This report focuses on performance testing of the ELVIS III device with new hybrid LVDTs provided by the Institute for Energy Technology (IFE), which is the world’s sole supplier of nuclear-grade LVDTs. The ELVIS III device was evaluated in terms of temperature showcasing promising results comparable to a typical type-K thermocouple (TC).

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2015

Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

316 Stainless Steel↗
DOE OSTI2015

Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to the qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Impacts Analyses Supporting the National Environmental Policy Act Environmental Assessment for the Resumption of Transient Testing Program

This document contains the analysis details and summary of analyses conducted to evaluate the environmental impacts for the Resumption of Transient Fuel and Materials Testing Program. It provides an assessment of the impacts for the two action alternatives being evaluated in the environmental assessment. These alternatives are (1) resumption of transient testing using the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) and (2) conducting transient testing using the Annular Core Research Reactor (ACRR) at Sandia National Laboratory in New Mexico (SNL/NM). Analyses are provided for radiologic emissions, other air emissions, soil contamination, and groundwater contamination that could occur (1) during normal operations, (2) as a result of accidents in one of the facilities, and (3) during transport. It does not include an assessment of the biotic, cultural resources, waste generation, or other impacts that could result from the resumption of transient testing. Analyses were conducted by technical professionals at INL and SNL/NM as noted throughout this report. The analyses are based on bounding radionuclide inventories, with the same inventories used for test materials by both alternatives and different inventories for the TREAT Reactor and ACRR. An upper value on the number of tests was assumed, with a test frequency determined by the realistic turn-around times required between experiments. The estimates provided for impacts during normal operations are based on historical emission rates and projected usage rates; therefore, they are bounding. Estimated doses for members of the public, collocated workers, and facility workers that could be incurred as a result of an accident are very conservative. They do not credit safety systems or administrative procedures (such as evacuation plans or use of personal protective equipment) that could be used to limit worker doses. Doses estimated for transportation are conservative and are based on transport of the bounding radiologic inventory that will be contained in any given test. The transportation analysis assumes all transports will contain the bounding inventory.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Comparison of animal and human blood for in vitro dynamic thrombogenicity testing of biomaterials

Abstract Background To determine suitable alternatives to human blood for in vitro dynamic thrombogenicity testing of biomaterials, four different animal blood sources (ovine, bovine, and porcine blood from live donors, and abattoir porcine blood) were compared to fresh human blood. Methods To account for blood coagulability differences between individual donors and species, each blood pool was heparinized to a donor‐specific concentration immediately before testing in a dynamic flow loop system. The target heparin level was established using a static thrombosis pre‐test. For dynamic testing, whole blood was recirculated at room temperature for 1 h at 200 ml/min through a flow loop containing a single test material. Four materials with varying thrombotic potentials were investigated: latex (positive control), polytetrafluoroethylene (PTFE) (negative control), silicone (intermediate thrombotic potential), and high‐density polyethylene (HDPE) (historically thromboresistant). Thrombus weight and surface area coverage on the test materials were quantified, along with platelet count reduction in the blood. Results While donor‐specific heparin levels varied substantially from 0.6 U/ml to 7.0 U/ml among the different blood sources, each source was able to differentiate between the thrombogenic latex and the thromboresistant PTFE and HDPE materials ( p < 0.05). However, only donor ovine and bovine blood were sensitive enough to differentiate an increased response for the intermediate thrombotic silicone material compared to PTFE and HDPE. Conclusions These results demonstrated that multiple animal blood sources (particularly donor ovine and bovine blood) may be suitable alternatives to fresh human blood for dynamic thrombogenicity testing when appropriate control materials and donor‐specific anticoagulation levels are used.

Engineering↗
DOE OSTI2015

Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence

In the framework of the development by CEA and SCK.CEN of a Fabry Perot Sensor (FPS) able to measure dimensional changes in Material Testing Reactor (MTR), the first goal of the SAKE 1 (Smirnof extention - Additional Key-tests on Elongation of glass fibres) irradiation was to measure the linear compaction of single mode fibres under high fast neutron fluence. Indeed, the compaction of the fibre which forms one side of the Fabry Perot cavity, may in particular cause a noticeable measurement error. An accurate quantification of this effect is then required to predict the radiation-induced drift and optimize the sensor design. To achieve this, an innovative approach was used. Approximately seventy uncoated fibre tips (length: 30 to 50 mm) have been prepared from several different fibre samples and were installed in the SCK.CEN BR2 reactor (Mol Belgium). After 22 days of irradiation a total fast (E > 1 MeV) fluence of 3 to 5x10{sup 19} n{sub fast}/cm{sup 2}, depending on the sample location, was accumulated. The temperature during irradiation was 291 deg. C, which is not far from the condition of the intended FPS use. A precise measurement of each fibre tip length was made before the irradiation and compared to the post irradiation measurement highlighting a decrease of the fibres' length corresponding to about 0.25% of linear compaction. The amplitude of the changes is independent of the capsule, which could mean that the compaction effect saturates even at the lowest considered fluence. In the prospect of performing distributed temperature measurement in MTR, several fibre Bragg gratings written using a femtosecond laser have been also irradiated. All the gratings were written in radiation hardened fibres, and underwent an additional treatment with a procedure enhancing their resistance to ionizing radiations. A special mounting made it possible to test the reflection and the transmission of the gratings on fibre samples cut down to 30 to 50 mm. The comparison of measurements made before and after the irradiation, at the same temperature, allowed us to measure the loss in reflectivity as well as the Bragg wavelength drift. The results are quite promising for some of the investigated gratings. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction are essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Distinguishing Research Reactor Fuels Using Neutron Multiplicity - 25615

Neutron measurements are one of the reliable means of interrogation of fissile material. Neutron measurements are used for total counting (pure metal fissile samples) or counting coincidence/multiplicity (for impure fissile samples). Various flavors of neutron measurement are used routinely by the IAEA to determine the fissile mass of an unknown sample. The objective of this work is to design an experiment to distinguish the two different fuel types using neutron multiplicity method. The design of experiment is performed using Monte Carlo N- Particle (MCNP) transport code. General Electric (GE) designed 1-MWt pool-type research reactor at UMass Lowell (UMLRR) uses Materials Testing Reactor (MTR) plate-type fuel consisting of 19 full and 2 partial fuel assemblies, controlled by four boron control blades and one regulating control rod. Two different fuel types are available at the UMLRR facility. In 2008, UMass Lowell obtained slightly burnt fuel assemblies from the decommissioned WPI research reactor. The simulation results discussed in this paper show that within the limits of statistical error, the single and double moments for the WPI and UMLRR fuels can be measured. The two fuels have very similar composition and enrichments resulting in very similar neutron multiplicity moments. This implies that neutron multiplicity measurements alone would not be sufficient to distinguish between the two fuel types at low burnups. The change in the observed single-double ratio between each fuel type is less than 6 % which falls below the uncertainty in the efficiency of the JCC71 slab detectors. The neutron signature from low burnup and low enriched uranium fuel are not strong, however, these can be measured under low background and long measurement times. This would be an important consideration in designing the experiments and parallel gamma measurements will also be planned. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Quantifying and Addressing the DOE Material Reactivity Requirements with Analysis and Testing of Hydrogen Storage Materials & Systems

The objective of this project is to examine safety aspects of candidate hydrogen storage materials and systems being developed in the DOE Hydrogen Program. As a result of this effort, the general DOE safety target will be given useful meaning by establishing a link between the characteristics of new storage materials and the satisfaction of safety criteria. This will be accomplished through the development and application of formal risk analysis methods, standardized materials testing, chemical reactivity characterization, novel risk mitigation approaches and subscale system demonstration. The project also will collaborate with other DOE and international activities in materials based hydrogen storage safety to provide a larger, highly coordinated effort.

08 HYDROGEN↗
DOE OSTI2025

Assessing the viability of a new subsize tensile specimen geometry for evaluation of structural nuclear and additively manufactured materials

The use of subsize specimens in nuclear materials testing has been a subject of ongoing interest due to radiation safety concerns and resource conservation needs. It is also of interest in additive manufacturing (AM), also known as 3D-printing, as subsize specimens can more accurately represent the behavior of the small, intricate geometries often produced using AM. A novel, extremely small geometry, called the Subsize Teeny (SST) was recently developed and is of interest for implementation. Given its extraordinarily small size and the complexities associated with subsize specimen testing, adequate vetting of this geometry is necessary to ensure data quality. Here, a variety of unirradiated nuclear structural materials were tested in the SST geometry and compared against the well-established SSJ3 geometry. In addition, two case studies implementing the SST as a screening geometry for AM materials were also conducted. The question of SST viability was found to be highly nuanced and will often be dependent on the context or application in question. It was determined, however, that the SST is a largely invalid geometry for exceptionally coarse-grained materials or in cases where the physical defect volume equals or exceeds 0.1 % or where the specimen machining parameters result in significant surface alterations. On the other hand, it was determined that the SST may be employed with confidence if the test material is nearly or totally free of physical defects, isotropic, demonstrates homogeneous plastic deformation, and possesses a fine-grained, nearly or totally homogeneous microstructure with at least twelve slip systems.

Additive manufacturing↗
DOE OSTI2014

1000–ton testing machine for cyclic fatigue tests of materials at liquid nitrogen temperatures

One of the main tasks of superconductive magnets R and D is to determine the mechanical and fatigue properties of structural materials and the critical design elements in the cryogenic temperature range. This paper describes a new facility built based on the industrial 1000-ton (10 MN) testing machine Schenk PC10.0S. Special equipment was developed to provide the mechanical and cyclic tensile fatigue tests of large-scale samples at the liquid nitrogen temperature and in a given load range. The main feature of the developed testing machine is the cryostat, in which the device converting a standard compression force of the testing machine to the tensile force affected at the test object is placed. The control system provides the remote control of the test and obtaining, processing and presentation of test data. As an example of the testing machine operation the test program and test results of the cyclic tensile fatigue tests of fullscale helium inlet sample of the PF1 coil ITER are presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2011

Mechanical Properties of Laser Heat Treated 6 mm Thick UHSS-Steel

In this work abrasion resistant (AR) steel with a sheet thickness of 6 mm was heat treated by a 4 kW Nd:YAG and a 4 kW Yb:Yag-laser, followed by self-quenching. In the delivered condition, test material blank (B27S) is water quenched from 920 deg. C. In this condition, fully martensitic microstructure provides excellent hardness of over 500 HB. The test material is referred to AR500 from now onwards. Laser heat treatment was carried out only on top surface of the AR500 sheet: the achieved maximum temperature in the cross-section varies as a function of the depth. Consequently, the microstructure and mechanical properties differ between the surfaces and the centre of the cross-section (layered microstructure). For better understanding, all layers were tested in tensile tests. For a wide heat treatment track, the laser beam was moved by scanning. Temperatures were measured using thermographic camera and thermocouples. Laser heat treated AR500 samples were tested in hardness tests and by air bending using a press brake machine. Microstructures were studied using a light microscope and FE-SEM/SEM-EBSD. At least three kind of microstructure layers were observed: 1) Dual-Phase ferritic/martensitic (T = A{sub C1}-A{sub C3}), 2) ferritic (T{approx}A{sub C3}) and 3) bainitic/martensitic (T>A{sub C3}).

36 MATERIALS SCIENCE↗
DOE OSTI2015

Design of smart 3D-digital X-ray microtomographic scanners for non-destructive testing of materials and components of electronic devices with a multilayered structure

The article studies the operating procedures of an X-ray microtomographic scanner and the module of reconstruction and analysis 3D-image of a test sample in particular. An algorithm for 3D-image reconstruction based on image shadow projections and mathematical methods of the processing are described. Chapter 1 describes the basic principles of X-ray tomography and general procedures of the device developed. Chapters 2 and 3 are devoted to the problem of resources saving by the system during the X-ray tomography procedure, which is achieved by preprocessing of the initial shadow projections. Preprocessing includes background noise removing from the images, which reduces the amount of shadow projections in general and increases the efficiency of the group shadow projections compression. Chapter 4 covers general procedures of defect search, which is based on vector analysis principles. In conclusion, the main applications of X-ray tomography are presented.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

Simple effective conservative treatment of uncertainty from sparse samples of random functions

This paper examines the variability of predicted responses when multiple stress-strain curves (reflecting variability from replicate material tests) are propagated through a finite element model of a ductile steel can being slowly crushed. Over 140 response quantities of interest (including displacements, stresses, strains, and calculated measures of material damage) are tracked in the simulations. Each response quantity’s behavior varies according to the particular stress-strain curves used for the materials in the model. We desire to estimate response variability when only a few stress-strain curve samples are available from material testing. Propagation of just a few samples will usually result in significantly underestimated response uncertainty relative to propagation of a much larger population that adequately samples the presiding random-function source. A simple classical statistical method, Tolerance Intervals, is tested for effectively treating sparse stress-strain curve data. The method is found to perform well on the highly nonlinear input-to-output response mappings and non-standard response distributions in the can-crush problem. The results and discussion in this paper support a proposition that the method will apply similarly well for other sparsely sampled random variable or function data, whether from experiments or models. Finally, the simple Tolerance Interval method is also demonstrated to be very economical.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Dissolved Air Flotation for Rapid Dewatering and Separation of Legacy Sludge Wastes - 19065

As a low footprint, high efficiency separation process, Dissolved Air Floatation (DAF) could effectively be retrofitted into existing waste management facilities at nuclear sites such Sellafield and Hanford to rapidly separate particulates from aqueous suspensions. The simplicity (no moving parts) and size of this technology coupled with low cost of construction and reagent purchase would also be ideal for ease of facility decommissioning with minimum impact to secondary waste generation. For this study, methyl isobutyl carbinol (MIBC) was used in this research as a frothing agent to produce a preferable stable foam. Mg(OH){sub 2} was selected as a test material as it is the result of long term corrosion of Magnox fuel in UK nuclear fuel storage ponds. Due to the cationic nature of the Mg(OH){sub 2} test material, the anionic surfactant, Sodium Dodecyl Sulphate (SDS) was used as collector to modify the surface properties to increase hydrophobicity. The performance of anionic SDS in floating 2.5%v/v Mg(OH){sub 2} was compared to the adsorption isotherm of SDS on Mg(OH){sub 2} to determine monolayer coverage surfactant dose which was investigated using Total Organic Carbon (TOC). SDS was found to increase particulate recovery to 93% with some water carry-over observed and it was found that particle bubble attachment was optimum for a select particle size distribution. This study proved that potential application of flotation as an efficient viable dewatering technique for common magnesium hydroxide based legacy wastes, using cheap, readily available collector and frother agents. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Genotoxicity of carbon nanofibers: Are they potentially more or less dangerous than carbon nanotubes or asbestos?

The production of carbon nanofibers and nanotubes (CNF/CNT) and their composite products is increasing globally. CNF are generating great interest in industrial sectors such as energy production and electronics, where alternative materials may have limited performance or are produced at a much higher cost. However, despite the increasing industrial use of carbon nanofibers, information on their potential adverse health effects is limited. In the current study, we examine the cytotoxic and genotoxic potential of carbon-based nanofibers (Pyrograf (registered) -III) and compare this material with the effects of asbestos fibers (crocidolite) or single-walled carbon nanotubes (SWCNT). The genotoxic effects in the lung fibroblast (V79) cell line were examined using two complementary assays: the comet assay and micronucleus (MN) test. In addition, we utilized fluorescence in situ hybridization to detect the chromatin pan-centromeric signals within the MN indicating their origin by aneugenic (chromosomal malsegregation) or clastogenic (chromosome breakage) mechanisms. Cytotoxicity tests revealed a concentration- and time-dependent loss of V79 cell viability after exposure to all tested materials in the following sequence: asbestos > CNF > SWCNT. Additionally, cellular uptake and generation of oxygen radicals was seen in the murine RAW264.7 macrophages following exposure to CNF or asbestos but not after administration of SWCNT. DNA damage and MN induction were found after exposure to all tested materials with the strongest effect seen for CNF. Finally, we demonstrated that CNF induced predominately centromere-positive MN in primary human small airway epithelial cells (SAEC) indicating aneugenic events. Further investigations are warranted to elucidate the possible mechanisms involved in CNF-induced genotoxicity.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2012

Thermal Analysis of Irradiation Experiments in the ATR

Reactor material testing in the INL's Advanced Test Reactor (ATR) involves modeling and simulation of each experiment to accurately determine the irradiation temperature. This paper describes thermal analysis of capsule experiments using gas gap temperature control and provides data on recent material tests that validate the modeling results. Static capsule experiments and lead-out capsule experiments are discussed. The source of temperature variation in capsule experiments and ways to mitigate these variations are also explained. Two examples of instrumented lead-out capsule experiments, TMIST-1 and UCSB-2, are presented. A comparison of measured and calculated temperatures is used to validate the thermal models and to ascertain the accuracy of the calculated temperature.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Experimental Study of RF Pulsed Heating

Cyclic thermal stresses produced by rf pulsed heating can be the limiting factor on the attainable reliable gradients for room temperature linear accelerators. This is especially true for structures that have complicated features for wakefield damping. These limits could be pushed higher by using special types of copper, copper alloys, or other conducting metals in constructing partial or complete accelerator structures. Here we present an experimental study aimed at determining the potential of these materials for tolerating cyclic thermal fatigue due to rf magnetic fields. A special cavity that has no electric field on the surface was employed in these studies. The cavity shape concentrates the magnetic field on one flat surface where the test material is placed. The materials tested in this study have included oxygen free electronic grade copper, copper zirconium, copper chromium, hot isostatically pressed copper, single crystal copper, electroplated copper, Glidcop(reg. sign), copper silver, and silver plated copper. The samples were exposed to different machining and heat treatment processes prior to rf processing. Each sample was tested to a peak pulsed heating temperature of approximately 110 C and remained at this temperature for approximately 10 x 10 6 rf pulses. In general, the results showed the possibility of pushing the gradient limits due to pulsed heating fatigue by the use of copper zirconium and copper chromium alloys.

36 MATERIALS SCIENCE↗