Results for “Nuclear cladding” — Page 12

DOE OSTI2016

Progress of GE development of accident tolerant fuel FeCrAl cladding

The Accident Tolerant Fuel (ATF) concept proposed by General Electric (GE) utilizes a FeCrAl alloy and current uranium dioxide (UO{sub 2}) fuel pellets, resulting in a fuel assembly that leverages the performance of existing LWR fuel assembly designs and infrastructure with improved accident tolerance. The focus of this paper is to report on the maturity of the concept and the overall feasibility on the use of ferritic FeCrAl alloys as cladding for nuclear fuel in commercial light water reactors. The following aspects are reviewed: mechanical properties, corrosion performance, stress corrosion cracking, fabrication assessments and performance evaluation. The results show that FeCrAl alloy clad fuel rods (with UO{sub 2} fuel) have reasonably demonstrated capabilities to meet or exceed current fuel design technical requirements while providing increased safety benefit during design basis events and severe accident conditions. Initial characterization of the materials, in particular APMT (the current leading FeCrAl alloy candidate whose composition is Fe + 22Cr + 5Al + 3Mo) indicates excellent operational and accident condition performance in both PWR and BWR environments. Challenges on the neutron economy and tritium release have a reasonable path to mitigation. The fabrication processes for the FeCrAl/UO{sub 2} system appears to be very similar to current LWR fuel fabrication processes (pilgering/extruding, heat treatments, welding, NDE techniques, etc.) which are mature and well understood. No issues are anticipated complying with current nuclear industry quality and performance standards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Post-quench ductility of zirconium alloy cladding materials - 2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperatures, followed by quenching in water, is of significant interest for the nuclear industry. High-temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents. The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility caused by differences in chemical composition, extensive testing on Zircaloy-4, ZIRLO{sup R}, and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation, and peak cladding temperature was performed. The results showed similar ductile-to-brittle transition for the alloys studied despite differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile to brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen-enriched layer that forms underneath the oxide, both during the time at high temperatures and during cooling. Significantly different morphologies of the oxygen-enriched layer were observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen-enriched layer are discussed. (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2020

Radiation-induced molecular hydrogen gas generation in the presence of aluminum alloy 1100

The United States government currently manages nearly 13 metric tons of aluminum-clad spent nuclear fuel (SNF) without a long-term storage solution, so a fundamental understanding of corrosion processes occurring on aluminum alloy surfaces is of utmost importance to plan for extended (>50 years) interim dry storage of aluminum-clad SNF. While thermal and chemical corrosion processes are well characterized for aluminum, radiation effects are not. To help understand the impacts of radiation on aluminum-clad SNF, here the radiation-induced molecular hydrogen gas (H 2 ) generation from pristine and pre-corroded aluminum alloy 1100 (Al-1100) coupons have been studied. Corrosion of coupons was achieved by submerging coupons in water at 95 °C for 29 days, yielding a ~5 μm boehmite/bayerite oxide film. Pre-corroded coupons were exposed to cobalt-60 gamma radiation to absorbed doses of up to 1.0 MGy under a variety of conditions: cover-gas composition (argon, nitrogen, or air), relative humidity (0, 50, and 100%), and temperature (ambient, 100, and 200 °C). Post-irradiation measurements demonstrated that the yield of H 2 was directly attributable to the presence of the Al-1100 coupons and their physisorbed water with dependence on absorbed gamma dose, relative humidity, and cover-gas composition. No H 2 was quantified in the presence of air, while both nitrogen and argon environments afforded higher H 2 yields with increasing relative humidities. This was attributed to the greater availability of adsorbed water for radiolytic processes. Irradiation of pre-corroded Al-1100 coupons at different temperatures under 0% relative humidity argon conditions yielded statistically equivalent H 2 yields for ambient temperatures and 100 °C. However, irradiation at 200 °C promoted a 3 to 4-fold increase in the yield of H 2 , possibly due to the transformation of bayerite to boehmite and/or improved efficiency of H • and H 2 release from oxide surfaces.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Post quench ductility of Zirconium alloy cladding materials - 6.2 STP-2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperature, followed by quenching in water, is of significant interest for the nuclear industry. High temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents (LOCA). The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility (PQD) caused by differences in chemical composition, extensive testing has been performed on Zircaloy-4, ZIRLO{sup R} and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation and peak cladding temperature. The results show similar ductile-to-brittle transition for the alloys studied, in spite of the differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile-to-brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen enriched layer that appears underneath the oxide, both during the time at high temperature and during cooling. Significantly different morphologies of the oxygen enriched layer are observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen enriched layer are discussed (authors)

36 MATERIALS SCIENCE↗

DOE OSTI2017

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Ion irradiation testing of Improved Accident Tolerant Cladding Materials

This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Applications of neutron radiography for the nuclear power industry

The World Conference on Neutron Radiography (WCNR) and International Topical Meeting on Neutron Radiography (ITMNR) series have been running over 35 years. The most recent event, ITMNR-8, focused on industrial applications and was the first time this series was hosted in China. In China, more than twenty new nuclear power plants are in construction and plans have been announced to increase the nuclear capacity further by a factor of three within fifteen years. There are additional prospects in many other nations. Neutron tests were vital during previous developments of materials and components for nuclear power applications, as reported in this conference series. For example a majority of the 140 papers in the Proceedings of the First WCNR are for the benefit of the nuclear power industry. Included are reviews of the diverse techniques being applied in Europe, Japan, the United States, and at many other centers. Many of those techniques are being utilized and advanced to the present time. Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Applications include examination of nuclear waste, nuclear fuels, cladding, control elements, and other critical components. In this paper, the techniques developed and applied internationally for the nuclear power industry since the earliest years are reviewed, and the question is asked whether neutron test techniques can be of value in development of the present and future generations of nuclear power plants world-wide.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Applications of Neutron Radiography for the Nuclear Power Industry

The World Conference on Neutron Radiography (WCNR) and International Topical Meeting on Neutron Radiography (ITMNR) series have been running over 35 years. The most recent event, ITMNR-8, focused on industrial applications and was the first time this series was hosted in China. In China, more than twenty new nuclear power plants are in construction and plans have been announced to increase the nuclear capacity further by a factor of three within fifteen years. There are additional prospects in many other nations. Neutron tests were vital during previous developments of materials and components for nuclear power applications, as reported in this conference series. For example a majority of the 140 papers in the Proceedings of the First WCNR are for the benefit of the nuclear power industry. Included are reviews of the diverse techniques being applied in Europe, Japan, the United States, and at many other centers. Many of those techniques are being utilized and advanced to the present time. Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Applications include examination of nuclear waste, nuclear fuels, cladding, control elements, and other critical components. In this paper, the techniques developed and applied internationally for the nuclear power industry since the earliest years are reviewed, and the question is asked whether neutron test techniques can be of value in development of the present and future generations of nuclear power plants world-wide.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Investigating Uranium Corrosion in Magnesium-containing Sludge Using X-ray Tomography - 20042

One of the current primary concerns within the British nuclear community is the decommissioning of legacy storage ponds in Sellafield Cumbria, where nuclear fuel cladding material (a Mg-Al alloy) resides along with fuel metallic swarf. Corrosion of the immersed metallic elements, which has been occurring over the storage period, has resulted in the formation of what is commonly known as Corroded Magnox Sludge. The behaviour of uranium metal within this particular environment and, more significantly, the identity of the corrosion products that may have formed is currently unknown. The present study attempts to shed some light on these aspects by investigating simplified surrogate systems, mimicking storage ponds. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Irradiation Vehicles for Evaluating SiC/SiC Cladding Bowing Under Neutron Flux Gradients

Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are among the most promising candidates for long term accident-tolerant nuclear fuel cladding. A key challenge related to their deployment is lateral bowing caused by differential radiation-induced swelling under dose or temperature gradients, which could obstruct coolant flow or interfere with control rod/blade movements. Although previous modeling efforts have predicted bowing behavior in light-water reactor (LWR) environments, experimental validation remains limited, especially at prototypic LWR temperatures. This study addresses that gap by irradiating six reduced-length SiC/SiC cladding tubes (~600 mm) in the High Flux Isotope Reactor (HFIR) at ~300°C, which is representative of LWR conditions. The tubes will be housed in a sealed vessel with an inert gas gap to maintain target temperatures and prevent direct coolant contact. Arranged in three pairs, each set will receive a different radiation dose (low, medium, high), with the central pair receiving ~0.1 displacements per atom (dpa)—the expected dose for peak bowing. The experiment will determine the dose-dependent bowing behavior and validate predictive finite element models. In this work, the tubes are freely suspended from pins to allow for unconstrained bowing; however, we present a concept for introducing localized constraints to represent grid spacer effects. Post-irradiation examination will include swelling measurements and profilometry to assess bowing and compare the results with model predictions. This work aims to confirm the conditions under which maximum bowing occurs so as to improve the reliability of SiC/SiC performance models in nuclear applications.

Russell, Nick [ORNL] (ORCID:0000000296099820)↗

DOE OSTI2012

The harmony between nuclear reactions and nuclear reactor structures and systems

Advanced nuclear energy is one extremely viable approach for achieving the required goals. With its extraordinarily high energy density (both, per unit mass and per unit volume), it produces over seven orders of magnitude less waste than fossil fuels per unit of energy generated. Applying nano-technologies to nuclear reactors could potentially produce the extraordinary performance required. The actual nuclear reactors lack of performances, the complexity and hazard of the fuel cycle are in part due to the lack of understanding of the nature's laws related to energy distribution applied to fission products, and in part to the current technologic capabilities that make the economical optimum. In order to produce the desired increase of performances a novel multi-scale multi-physics and engineering approach have been developed, starting from the nuclear reactions involved, analyzing in detail the key features and requirements of the 'key players' in the process (neutrons, compound nucleus, fission products, transmutation products, decay radiation), the consequences of their interaction with matter. That complex interaction generates new reactions and new key-players (knock-on electrons, photons, phonons) that further interact with the matter represented by the nuclear fuel, cladding, cooling agents, structural materials and control systems. The understanding of this complexity of problems from fm-ps scale up to macro-system and mitigating all the requirements drives to that desired harmony that provides a safe energy delivery. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Zircaloy Oxide Growth

The goal of the research is to create a standard for growing an oxide layer on zircaloy nuclear fuel cladding to aid in qualifying new eddy current probes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Impact of corrosion layer composition and thermal pretreatment on the radiolytic generation of molecular hydrogen from aluminum alloys

To fully evaluate the feasibility of extending the dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters, the amount of radiation-induced molecular hydrogen (H 2 ) generated and the impact of absorbed radiation dose on the ASNF corrosion layer composition must be accurately assessed. Here, in this study, we report a slowing in the rate of radiolytic H 2 generation for pre-corroded AA1100 and AA6061 aluminum alloy coupons irradiated to up to 53 MGy of absorbed cobalt-60 gamma dose. By exploring a variety of thermal pretreatment conditions for AA6061 coupons, we find that the “steady-state” yield of H 2 depends more on the aluminum alloy used than on the treatments applied prior to dry storage. Scanning electron microscopy and positron annihilation lifetime spectroscopy techniques also provided evidence for gamma radiation-induced defects in the corrosion layers of the investigated aluminum alloy coupons for high absorbed doses (~50 MGy), the consequences of which on cladding integrity and H 2 generation should be explored in future works.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗

DOE OSTI2013

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term and Elevated Temperature Irradiation: Modeling and Experimental Investigation

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation. This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime. Specifically, we propose four research thrusts, namely Thrust 1: Identify the formation mechanism and evolution for dislocation loops with Burgers vector of a<100> and determine whether the defect microstructure (predominately dislocation loop/dislocation density) saturates at high dose. Thrust 2: Identify whether a threshold irradiation temperature or dose exists for the nucleation of growing voids that mark the beginning of irradiation-induced swelling, and begin to probe the limits of thermal stability of the tempered Martensitic structure under irradiation. Thrust 3: Evaluate the stability of nanometer sized Y- Ti-O based oxide dispersion strengthened (ODS) particles at high fluence/temperature. Thrust 4: Evaluate the extent to which precipitates form and/or dissolve as a function of irradiation temperature and dose, and how these changes are driven by radiation induced segregation and microchemical evolutions and determined by the initial microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Microstructure and Property Evolution in Advanced Cladding and Duct Materials Under Long-Term Irradiation at Elevated Temperature: Critical Experiments

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by microchemistry changes due to radiation-induced segregation, dislocation loop formation and growth, radiation induced precipitation, destabilization of the existing precipitate structure, as well as the possibility for void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiation-induced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses to 200 dpa and beyond). Further, predictive modeling is not yet possible, as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. This project builds upon joint work at the proposing institutions, under a NERI-C program that is scheduled to end in September, to understand the effects of radiation on these important materials. The objective of this project is to conduct critical experiments to understand the evolution of microstructural and microchemical features (loops, voids, precipitates, and segregation) and mechanical properties (hardening and creep) under high temperature and full dose range radiation, including the effect of differences in the initial material composition and microstructure on the microstructural response, including key questions related to saturation of the microstructure at high doses and temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Multiphysics modeling of two-phase film boiling within porous corrosion deposits

Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits. Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys. - Highlights: • A two-phase model of CRUD's effects on fuel cladding is developed and improved. • This model eliminates the formerly erroneous assumption of wick boiling. • Higher fuel cladding temperatures are predicted when accounting for two-phase flow. • Double-peaks in thermal conductivity vs. heat flux in experiments are explained. • A “double dryout” mechanism in CRUD is proposed based on the model and experiments.

42 ENGINEERING↗

DOE OSTI2017

Volatility Studies of Lanthanide Hexafluoroacetylacetone Chelates

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy (DOE) Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors (LWRs). Some of the leading candidates to replace traditional zirconium-based cladding are alumina-forming ferritic alloys (e.g., FeCrAl) and silicon carbide (SiC) composites. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative geometry in the High Flux Isotope Reactor (HFIR) under relevant LWR temperatures and in some cases under prototypic heat flux. These designs allow for post-irradiation examination (PIE) of cladding which closely resembles expected commercially viable geometries and microstructures. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost for moving advanced cladding materials closer to commercialization. This work describes the capsule designs that have been developed at ORNL, some initial results, and plans for future irradiations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Evaluation of the ring tension test (RTT) for robust determination of material strengths

The ring tension test (RTT) is an experimental method for determining mechanical behavior in a material’s circumferential or hoop direction. It is a crucial test for testing anisotropic materials with tube geometry, such as nuclear fuel cladding or irradiated pipes. Several RTT configurations exist, each with their own advantages and disadvantages. However, this test is significantly more complex than traditional tensile testing and can be especially sensitive to small differences and inconsistencies in the test setup and geometry, ultimately affecting the derived mechanical properties. Previous research has focused on method development, and little work has been done on understanding the subtle differences between an ideal test and experiments, specifically when the tests are performed on highly irradiated materials in hot cells. In this work, a finite element-based investigation of the RTT is conducted to fill this gap. The two most used test configurations are investigated, comparing their ability to determine accurate material strengths through plastic deformation. Several non-ideal conditions and uncontrollable effects which are likely to occur during experimental testing such as machining tolerances, variations of specimen geometry from nominal dimensions, rotation of specimens and fixturing, and other test setup discrepancies are studied. The sensitivity of measured strengths to these conditions is presented. A mechanics-based approach to describing and correcting raw data to determine actual strengths is also included for one of the configurations, resulting in a robust correction method with highly accurate material strength measurements. Finally, based on these analyses, the hemicylindrical mandrel configuration is recommended with a gauge region oriented at a 45° angle.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗