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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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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Acid Dissolution Behavior of Ferritic FeCrAl Tubes Candidates for Nuclear Fuel Cladding

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel in light water reactors. One solution is to replace the zirconium cladding using ferritic iron-chromium-aluminum (FeCrAl) alloys, which offer extraordinary resistance to high-temperature reaction with air or steam due to the formation of a protective alumina layer on the external surface. It is important to characterize the behavior of FeCrAl not only during accident conditions but in the entire fuel cycle, which may include reprocessing of the used fuel after it is removed from the power reactors. The reprocessing may involve the dissolution of the fuel rods in mineral acids. Little or nothing is known on the dissolution of FeCrAl alloys in common mineral acids, therefore the objective of this research was to study the dissolution of typical cladding tubing having two compositions of FeCrAl (APMT and C26M) in three acids (H 2 SO 4 , HNO 3 , and HCl) as a function of the temperature using both standard ASTM immersion tests as well as electrochemical tests. The dissolution behavior of the FeCrAl alloys is compared to the dissolution capability of other traditional nuclear materials such as austenitic stainless steels (304SS and 316SS) and austenitic nickel alloys (Alloy 600 and Hastelloy C-276). Results show that both C26M and APMT have a higher dissolution capability in the studied mineral acids, which will be beneficial for reprocessing procedures.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Energy efficient soundproofing window retrofits

Described herein are window retrofits including a monolithic silica aerogel slab having (i) an average haze value of <5% as calculated in accordance with ASTM standard D1003-13 and (ii) a U-factor of <0.5 BTU/sf/hr/° F., and a transparent polymer envelope sealed at an internal pressure of ≤1 atmosphere, wherein the monolithic silica aerogel slab is encapsulated in the transparent polymer envelope. The monolithic aerogel slab can have a transmittance >94% at 8 mm thickness. The window retrofit can be bonded to a glass sheet.

Wang, Evelyn N.↗
DOE OSTI2022

Advanced Graphite Creep Uncertainty Analysis

Radiation damage estimation is an important component of the post irradiation analysis of the Advanced Graphite Creep (AGC) experiment. It depends primarily on the fast fluence, which is determined using well established methods of spectral adjustment. These are based on best estimates from models such as Monte Carlo N-Particle (MCNP), input cross-sections, and measured activities from flux wires in the experiment. Each of these parameters can propagate uncertainties which will affect the uncertainty in the calculated dose levels for AGC, or any experiment irradiated within a reactor. While the methods of propagating uncertainty are well-established, the final uncertainty estimates they provide are only as good as the estimates of uncertainty in the inputs on which they are based. The purpose of this work is to outline some deficiencies in the ways these input uncertainties are presently estimated, and to outline a methodology by which they can be improved. The fast fluence and radiation damage received by graphite specimens irradiated in the Advanced Graphite Creep (AGC) experiments is presently estimated using spectral adjustment methods that are based on both flux wire activity measurements, and MCNP model predictions. This work describes an ongoing effort to quantify and propagate uncertainties in inputs to the spectral adjustment process, and thereby quantify the resultant error in radiation damage (dpa) estimates. The effort is multi-faceted, and we consider the impacts of both the set of flux wires selected, and the counting process. An expanded set of flux wires is identified that provides a more comprehensive data set on the fast spectrum. To address the counting process itself, a series of round-robin measurements in several reactor metrology laboratories across the Department of Energy (DOE) complex and nuclear industry are being undertaken to refine the American Society for Testing and Materials (ASTM) standards for flux wire measurements. To address the contribution of uncertainty in the MCNP model predictions, an uncertainty quantification (UQ) tool has been developed that statistically samples the model input parameters, runs a series of cases, and assimilates the results to provide an overall uncertainty. The impact of the MCNP UQ tool results is demonstrated by re-analyzing previous AGC flux wire and irradiation data. While the expanded flux wire set obviously cannot be added to these experiments retroactively, plans for future graphite irradiations are outlined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

316L Stainless Steel Wire Arc Additive Manufacturing and Characterization for Potential SNF Canister Production

In this paper, 316L stainless steel (316L SS) straight walls, curved walls, and vessels were printed by wire arc additive manufacturing (WAAM) using the 316L SS welding wire, to demonstrate the feasibility of spent nuclear fuel (SNF) canisters using this advanced manufacturing technique. Helium leak test with leak tight criteria of 1 × 10-7 ref-cc/s were performed on printed vessels with various wall thickness, and testing pressures were from 0.345 MPa (50 psi) to 1.724 MPa (250 psi). Printed wall microstructures were characterized using optical microscopy (OM) and scanning electron microscopy (SEM). Tensile specimens were machined out from printed walls along the length, height, and thickness directions, respectively, and tested at different temperature conditions, room temperature, 80 °C, 150 °C and 250 °C, respectively. Results showed t hat, all five WAAM-printed canisters with different thickness passed the helium leaktight criteria of radioactive materials containers by the ANSI N14.5; Dispersed micrometer-level size inclusions/voids were observed at printed wall cross sections, dendritic microstructures were found with grain size variation through the welding torch shifting direction, and very small amount of δ-ferrite and σ-phase were also observed in those cross sections; The printed walls were ductile, all tensile specimens broke with large plastic deformation/elongation and without brittle failure, and printed specimens yield strengths and tensile strengths under room temperature and 80 °C conditions are higher than minimum values of 316L SS for pressure vessels and for general applications at room temperature required by the ASTM standard, respectively.

Tang, Wei↗
DOE OSTI2023

Additively Manufactured Strain Sensing for Nuclear Reactor Applications

Real-time monitoring of materials in harsh environments is a crucial technique towards reducing innovation time in nuclear systems. The successful measurement of real time, in-situ strain measurements during nuclear reactor operation requires innovative sensing solutions, including novel sensing strategies as well as advanced sensor design, manufacturing, and materials selection. In this paper, we will discuss two primary strategies for in-situ strain measurement strategies: additively manufactured (AM) capacitive strain gauges (CSGs) and digital image correlation (DIC). Current commercial strain gauges have limited applications in reactors due to the harsh operating conditions and non-trivial attachment strategies (i.e., welding, epoxy-adhesive) that can affect both the sensing performance and the underlying substrate under testing. AM CSGs are a viable solution as they have a low profile, low hysteresis, and wireless sensing integration capabilities that will enhance nuclear sensing technologies. In this work, the mechanical and thermal performance of the AM CSGs were tested up to 300 °C using ASTM standardized testing procedures to simulate the temperatures found in existing light water reactors. The AM CSGs had a similar performance across multiple samples which correlates to analytical models. This work leads towards the development of CSGs designed for higher temperatures and additional environmental factors found in Generation-IV reactors. Non-contact sensors, such as DIC, offer a less destructive way to measure deformation of materials when compared to alternative methods of in-situ strain determination, such as weldable strain gauges. However, DIC requires high contrast surfaces, which often relies on the implementation of artificial patterns. Using traditional splatter techniques to fabricate these patterns have limitations, including poor surface adhesion and reproducibility. In this work, AM fabrication techniques were implemented to avoid such limitations. Accordingly, aerosol jet printing (AJP) was used to print small scale periodic patterns of silver on stainless steel and aluminum tensile specimens. DIC was employed to monitor strain (up to 1100 µe) during temperature cycling from 23-600 °C. The results validated the use of AJP to better control pattern parameters for small fields of view applications at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Advancement of Commercial Intumescent Expanding Foams for Deactivation and Decommissioning in the Nuclear Sector - 20198

Florida International University (FIU), in collaboration with The Department of Energy's Office of Environmental Management (DOE-EM), Savannah River National Laboratory (SRNL), and sites across the Savannah River complex, have identified an operational requirement for a fixative technology that is intended to immobilize and/or isolate residual contamination within a 3-dimensional space. Fixation of radiological contamination can reduce worker risk and mitigate potentially hazardous conditions, however nearly every marketed contamination fixative has been found to be flammable; a significant concern in radiological facilities. Coupled with this, industry fixatives are normally used as a thin coating which can present problems when attempting to stabilize irregular geometry or areas that are difficult to access whilst ensuring full coverage. The technical evaluation and advancement of commercial-off-the-shelf (COTS) polyurethane foams has yielded a down-selected candidate that shows potential in meeting the requirements to support deactivation and decommissioning activities. Several performance criteria have been established and tested to progress the technology readiness level towards an active field demonstration (TRL-7). Such criteria include: mechanical failure limits, adhesive and cohesive properties, thermal/fire resilience, determining thermal behavior, ability to immobilize contamination, and a means of non-destructive evaluation of applications. The test scenario examined was targeted towards an application for decommissioning nuclear pipework, in which the down-selected polyurethane foam would act as a barrier to segregate pipework and mitigate the potential for release during cutting, packaging, and storage operations. Testing carried out at SRNL included: mechanical evaluation of tensile, compressive, and adhesion strength by dynamic mechanical analysis (DMA), as well as thermogravimetric analysis (TGA). FIU examined the foam's fixative properties by utilizing phosphorescent europium-dysprosium doped strontium aluminate powder to investigating the extent to which contamination can be immobilized. FIU has also exploited previous successes in the field of intumescent technologies to assess the down-selected foam's tolerance to an extreme fire scenario, while maintaining the ability to effectively mitigate a contamination release. Parallel to this, extensive thermal investigations were carried out to determine the upper boundary of anticipated heat generation during the curing process as heat generation has the potential to compromise rubber parts of contaminated enclosures. These investigations subsequently yielded a promising method for a non-destructive application evaluation by means of infrared thermography. Utilizing the high sensitivity of modern IR cameras, coupled with the heat generated during the curing process of the polyurethane foam, FIU has been exploring the concept of monitoring the external pipe surface for indications of an irregular or abnormal application, thus informing operational decision making. The testing carried out utilized several current 'best fit' ASTM standards, which serve as helpful guidelines for testing, however, a precise definition of the operational parameters and requirements is still necessary. With continued collaboration with SRNL, FIU aims refine said definitions and develop new standards by which this, and other decommissioning technologies, can be accredited by relevant standards based testing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Physical Properties of Hilti HIT-RE 500 V3 After Irradiation: Slant Shear, Tensile, Flexural, and Compressive

There are various applications in which epoxy adhesives may be used within hot-cell environments. These epoxies currently see limited use within hot cells since it is known that organics see degradation after small doses even as low as 50 kilogray (kGy). These losses in physical property strengths limit the use of these epoxies to non-structural applications. In modern construction, however, the use of epoxies to bond structural components such as rebar or threaded anchors is becoming increasingly prevalent. To employ these types of materials the Idaho National Laboratory (INL) has begun investigating the specific degradation properties of epoxies for use in structural applications in environments where radiation is present. Hilti HIT-RE 500 V3 is used extensively in new construction and renovation projects across the world. The INL has conducted various tests on this epoxy for low dose environments. These tests include tensile, flexural, compression, and slant shear tests following the American Society for Testing Material International (ASTM) standards D638, D790, D695, and C882/882M respectively. Two sample sets for each test were created where the baseline specimens were not irradiated, and another set was irradiated to an accumulated dose of approximately 50-60 kGy. All the samples were destructively tested and analyzed focusing on the change in physical properties, examples of these tests are shown in Figures 1-4. Tensile testing results indicate a 0.9% loss in maximum load capacity, a 0.5% loss in tensile stress at maximum load, a 1.4% loss in tensile strain at break, and a 15% decrease in modulus of elasticity. Flexural testing results display an increase in all the following: 12% in strain at maximum load, 10% in maximum load capacity, 5% in stress at maximum load, and 3% in young’s modulus. The compressive physical properties after irradiation display a 1% increase in maximum load, a 3% decrease in stress at yield, and a 5% decrease in modulus of elasticity. The maximum compressive stress saw no measurable change after irradiation. The slant shear specimens after irradiation displayed a 2% increase in maximum load capacity and maximum compressive stress. Though a structural analysis would have to be completed as is normal for building design the increases and decreases in the measured physical properties indicate that this material may be used for structural applications in low dose hot-cell environments without adhesive failure.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Elevated Temperature Graphite Mechanical Testing

High purity graphite will be used for core components within most High Temperature Reactor (HTR) designs. Several "high tech” industries currently utilize synthetic, high-purity, commercially available graphite components to fabricate photovoltaic cells, semi-conductors, optical fibers, and other high value electronic industry products. New advanced HTR designs are also interested in using these graphite grades for long-term, internal core component applications. Consequently, the US Department of Energy, Advanced Reactor Technologies (DOE-ART) program has spent several years testing different high purity graphite grades for potential use within these new nuclear reactor designs. A significant part of that effort has been in the development and improvement of American Society for Testing and Materials (ASTM) test standards specifically for nuclear graphite grades. Nearly all ASTM test standards have either been developed or improved by the DOE-ART program over the past 25 years. The ART program continues to assist in the development of new ASTM test standards in support of the future commercial HTR fleet currently being designed and built in the USA. The newest effort undertaken by ART is the development of high temperature mechanical testing practices that may be acceptable for a future ASTM test standard (or guide) for this critical material property measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Influence of fatigue precracking and specimen size on Master Curve fracture toughness measurements of EUROFER97 and F82H steels

EUROFER97 and F82H steels are two leading reduced-activation ferritic-martensitic (RAFM) steels for fusion first wall and blanket applications. Exposure to the harsh environment of fusion reactors can result in severe degradation of fracture toughness. Thus, the post-irradiation evaluation of fracture toughness is critical for understanding the material behavior. Due to the space constraints of irradiation facilities and challenges in controlling a uniform irradiation condition for large size specimens, the development of small specimen test techniques (SSTT) is indispensable to evaluate the performance of irradiated materials. In this study, we evaluated specimen size effects on the Master Curve fracture toughness of EUROFER97 and F82H steels. A wide variety of specimens, including 0.5 T compact tension (C(T)) specimens, 0.16 T mini-compact tension (miniC(T)) specimens, and 1.65 mm miniature bend bar specimens, were tested. The testing methodology was based on the Master Curve method in the ASTM E1921 standard. No specimen size effect was observed in 0.5 T C(T) and 0.16 T miniC(T) specimens on the Master Curve reference temperature T0, while 1.65 mm miniature bend bar specimens yielded a higher T 0Q . A strong effect of fatigue precracking on T 0 for 0.5 T C(T) and 0.16 T miniC(T) specimens was observed, such that testing on specimens with skewed fatigue precrack fronts resulted in lower T 0 than for specimens with ASTM standard qualified straight fatigue precrack fronts. The results highlight the importance of experimental quality control in developing SSTT for Master Curve fracture toughness testing. Lastly, we also evaluated and provided recommendations on the minimum number of specimens needed for each specimen type for yielding reliable T 0Q values.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Mechanical behavior analysis using small-size graphite disc compression testing with digital image correlation methods

The ASTM International Standard Test Method for Tensile Stress-Strain of Carbon and Graphite (ASTM C749) requires dog-bone shaped specimens. Unfortunately, specimen sizes required in ASTM C749 are much too large for irradiation within a test reactor and are incompatible with irradiation capsule volumes, molten salt degradation facilities, or oxidation apparatus. The ASTM International Standard Test Method for Tensile Strength Estimate by Disc Compression of Manufactured Graphite (ASTM D8289) was developed as an alternative that provides a means for testing tensile properties using smaller specimens when geometric constraints are present. However, the important stress–strain relation throughout the test loading history that could be recorded using ASTM C749 is lost by using ASTM D8289 because of the difficulties involved with using traditional strain gauges or extensometers on such a small specimen. In this work, a digital image correlation (DIC) method, which uses a full-field noncontact surface displacement measurement technique, was applied along with the ASTM D8289 splitting tensile test on eight small samples of graphite grade Mersen 2114 to track their deformation during compressive loading. Initial results showed that the DIC technique can measure the surface displacement/strain on these small (Ø6 × 3 mm) graphite specimens with high accuracy and good repeatability, indicating that the DIC technique, combined with the small-disc splitting tensile test in ASTM D8289, can capture the strain evolution in the test. In conclusion, further comparison of measured strains with analytical and numerical simulation results provides additional understanding of the mechanical behaviour of this nuclear graphite material in complex stress states.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Investigation of nonlinear mechanical behavior of two superfine-grained graphites with DIC assisted disc splitting test

Using standardized uniaxial tensile test specimens is not practical due to the limited volumes in irradiation capsules, molten salt degradation facilities, or oxidation apparatus. The ASTM International Standard Test Method for Tensile Strength Estimate by Disc Compression of Manufactured Graphite (ASTM D8289) was developed to provide a convenient way for estimating tensile strength. Unlike the traditional uniaxial tensile test in ASTM C749 (International Standard Test Method for Tensile Stress–Strain of Carbon and Graphite), which uses dog-bone shaped specimens larger than 12.95 mm × 120.65 mm, the ASTM D8289 standard uses smaller discs with diameters of 6–12.7 mm. A digital image correlation (DIC) system, which uses a full-field noncontact surface displacement measurement technique, was applied along with the ASTM D8289 disc splitting test on IG-110 samples and compared with previous measurements from Mersen 2114 samples. Results confirmed that the DIC technique can measure the surface displacement/strain on these small (Ø6 mm × 3 mm) graphite specimens with good repeatability. However, Mersen 2114 and IG-110 samples exhibited strain discrepancies when DIC measurements were compared with analytical and finite element simulation values. The loading history and strain results also indicated different mechanical behaviors between Mersen 2114 and IG-110, particularly the nonlinear behavior of the IG-110 samples. Good agreement was observed by comparing the splitting tensile strengths of two superfine-grained grades with results from other work. The specimen size effect is discussed when comparing the splitting tensile strength with corresponding uniaxial tensile strength of these two graphite grades.

Lin, Lianshan [ORNL] (ORCID:0000000203399219)↗
DOE OSTI2024

Effect of Sample Thickness on the Tensile Strength of Small Graphite Discs

This report formally documents the completion of the Advanced Reactor Technologies (ART) Level 3 Milestone (M3TG-24OR0501054), “Continue activities related to Split Disc-DIC - complete analysis of effect of sample thickness on one fine grain graphite,” due May 31, 2024. Details within this report outline the status of activities aimed at elucidating the effect of sample thickness on the tensile strength measurement using the small-disc, split-disc testing standard in support of the US Department of Energy’s ART Graphite R&D Program. The ASTM D8289, Standard Test Method for Tensile Strength Estimate by Disc Compression of Manufactured Graphite, was developed to provide an alternative means for testing tensile strength on smaller specimens, which are compatible with available irradiation capsule volumes, and other compatibility measurements. ASTM D8289 specifies that acceptable specimen diameter can range from 6 to 12.7 mm and that the maximum allowed sample thickness should be half of the diameter. However, information is limited with respect to the effect of thickness on the measured splitting tensile strength. This report documents efforts to understand the effect of sample thickness on splitting tensile strength. The work involved testing Ø12.7 mm samples of fine-grain graphites 2114 and IG-110 of different thicknesses (6.35, 5, 4, and 3 mm). The digital image correlation (DIC) method was applied to the samples, along with the ASTM D8289 Standard, to help interpret the measured results.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Determination of Water Content in Bio-Oils by Volumetric Karl Fischer Titration: Laboratory Analytical Procedure (LAP)

The water content of bio-oils is a key metric for several reasons. Water is typically the most concentrated single component of fast pyrolysis oils, though this will be much lower in catalytic fast pyrolysis and upgraded products. Reduced water content is preferable in bio-oil as water can contribute to phase separation, corrosivity, and instability. Additionally, lower water is beneficial for physical properties such as energy density. The percentage of water can also be used to correct the calculation for organic oxygen when determining total oxygen content of bio-oils via combustion-based ultimate analysis. The procedure described here is specifically written for the analysis of bio-oils, and is based on ASTM E203, Standard Test Method for Water Using Volumetric Karl Fischer Titration. This standard test method is specified for measuring the mass % water in pyrolysis liquid biofuels in ASTM D7544, Standard Specification for Pyrolysis Liquid Biofuel. Although the standard method is prescribed for the analysis of pyrolysis liquids for use in industrial and commercial burners, the method is not specifically written for this product. The method described with this Laboratory Analytical Procedure (LAP) provides specific guidance for the analysis of bio-oils. Both organic and aqueous phases can be measured with this technique.

09 BIOMASS FUELS↗
DOE OSTI2026

Do Flashing and Sealant Materials Need to be Exposed to Extreme Cycling?

In the U.S., over 50% of homes were built before 1980, when minimal building codes existed. The aging of building envelopes can significantly contribute to the air leakage, moisture intrusion, and durability of buildings leading to higher energy demands. Although the durability of building materials and envelope systems has been studied extensively, the durability of specific components like window-wall flashing and sealants remains underexplored. Current standards such as ASTM, NIST, and AAMA primarily focus on durability testing considering the high temperatures only and require sealant testing on anodized aluminum panels, leaving questions about performance in cold or fluctuating temperatures unanswered. Current ASTM and AAMA standards provide guidelines for durability testing but predominantly focus on exposure to higher temperature conditions. It remains unclear how exposure to cold temperatures, or cycles of both hot and cold conditions, affect the durability of these materials. Furthermore, AAMA standards mandate that flashing materials be tested on anodized aluminum panels, which may not reflect real-world applications. Therefore, this initial study aims to explore this concern through testing by constructing two window-wall interface prototypes and applying five different flashing and sealant materials to their joints. These samples were subjected to extreme temperatures, ranging from −34 ℃ (−30 °F) to 50 ℃ (122 °F), over a 28-day period (fourteen cycles of 24 hours), with daily visual inspections to document any signs of deterioration such as wrinkling, distortion, or warping. The study has two main objectives: (1) to determine whether flashing and sealant materials lose their durability and strength when exposed to varying climate conditions, specifically cold climates, and (2) to assess whether the interaction of these materials with different substrates (e.g., lumber, PVC, insulation) affects their durability and strength under varying climate conditions. The findings demonstrate how aging impacts the durability of window-wall systems interfaces, justifying future research that potentially could lead to the development of new testing protocols if significant defects are identified.

Palani, Hevar [ORNL] (ORCID:0000000220211994)↗
DOE OSTI2021

Workshop on Using a Git-based Repository for ASTM Committee Technical Knowledge Capture [Slides]

Consensus technical standards maintained by ASTM International rely on individual contributors to act as technical contacts, effectively stewards, for a particular standard. As a particular standard’s technical content evolves over time, the contact must update relevant tables, graphics, and supporting discussion. However, the details of the data sources and/or processes used to perform such updates may be separate enough from the standard such that expert knowledge is required to perform these updates. As a result of this need for expert knowledge, and the desire to capture that knowledge, subcommittee E10.05 (Nuclear Radiation Metrology) has experimented with using a git-based repository to capture data and processes to reproduce figures needed by E10.05 to maintain the ASTM standards under their jurisdiction. This workshop will provide an overview of the git-based repository site’s organization, operation, and outputs. It will then lead interested workshop participants through adding content, not yet already included in the repository to the site.

42 ENGINEERING↗
DOE OSTI2020

Comparison of fluence-rate determination techniques for the Advanced Test Reactor-Critical Facility

A technique using small, enriched uranium-aluminum alloy monitors to measure fission rates in fuel plate assembly reactors was developed at the Idaho National Laboratory Radiation Measurements Laboratory in support of Advanced Test Reactor Critical (ATR-C) Facility irradiations. Previous measurements were performed to determine the correlation between the beta-particle activity and the fission rate in the U-Al monitors following irradiation. The American Society for Testing and Materials (ASTM) has established a standard practice E704 for determining neutron fluence rate by radioactivation of depleted uranium by measuring the concentration of fission-products in irradiated uranium monitors. Rather than measuring the beta-particle decay from fission-products, the decay rate of fission-products is determined by gamma-ray spectroscopy. The feasibility of using a method adapted from the ASTM standard practice E704 was investigated for ATR-C irradiations and compared to the present method by analyzing a subset of 21 monitor wires using both methods. The resulting calculated fission rates agreed with an average standard deviation of 7.45%, within the average combined measurement error of 9.05%. Practical considerations such as expected fluence-rates, sample activity, and counting-time were compared between the two methods and the sources of error for each method were investigated. Quality assurance measurements utilizing the modified ASTM standard method are recommended for future test irradiation, but gross-beta-particle counting remains the most practical solution for ATR-C fluence-rate determination.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Evaluation of the GaAs displacement damage metric using updated nuclear data

The emerging use of the physics-based athermal recombination-corrected displacement per atom (arc-dpa) model for the displacement damage efficiency has motivated a re-evaluation of the historical empirically-derived GaAs damage response function with the purpose of highlighting needs for future analytical and experimental work. The 1-MeV neutron damage equivalence methodology used in the ASTM E-722 standard for GaAs has been re-evaluated using updated nuclear data. This yielded a higher fidelity representation of the GaAs displacement kerma and, through the use of the refined PKA recoil energy-dependent damage efficiency model, an updated 1-MeV(GaAs) displacement damage function. This re-evaluation included use of the Norgett-Robinson-Torrens (NRT) model for an updated threshold treatment, rather than the sharp-threshold Kinchin-Pease model used in the current ASTM standard. The underlying nuclear data evaluations have been updated to use the ENDF/VIII.0 {sup 75}As and TENDL-2019 {sup 71}Ga/{sup 69}Ga evaluations. The displacement kerma and 1-MeV-equivalent damage responses were calculated using a modified NJOY-2016 code which allowed for refinements in some of the damage models. This paper shows that an updated displacement damage function, based upon the latest nuclear data, is consistent with the experimental data used to develop the current ASTM E-722 GaAs standard. Using a double ratio approach to compare the available experimental data with the calculated response, the average legacy double ratio was found to be 0.97 ± 0.05 and the average updated double ratio was found to be 0.94 ± 0.05. (authors)

36 MATERIALS SCIENCE↗