Search NASA⌕ Search

SEARCH · Search NASA

Results for “TEST REACTOR”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

University of Missouri Research Reactor LEU Fuel Element Flow Test Conceptual Design—Hydraulic Reactor Design Parameters

The University of Missouri-Columbia Research Reactor (MURR®) is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that actively collaborates with the National Nuclear Security Administration (NNSA) Material Management and Minimization(M 3 ) Reactor Conversion Program to convert to the use of low-enriched uranium (LEU, < 20 wt.% U-235) fuel. A new type of LEU fuel with very high density, based on an alloy of uranium and 10 weight percent molybdenum (U-10Mo), is expected to allow the conversion to LEU of USHPRR that have been found unable to be converted with previously qualified uranium silicide-aluminum (U 3 Si 2 -Al) dispersion fuel. MURR has been working with the USHPRR Reactor Conversion (RC) Pillar at Argonne National Laboratory to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analyses, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary Safety Analysis Report (SAR) for conversion to LEU fuel. This work is performed in preparation for the flow test campaign that will be conducted by the USHPRR RC Pillar. The purpose of the hydraulic performance evaluation of the MURR LEU fuel element designed by the RC Pillar is to test a prototypic commercially fabricated LEU fuel element to determine whether any failure modes are observed or predicted in the fuel element, including significant deformations such as plate bending, twisting, or plate detachment from the side plate under selected safety-basis limits for reactor hydraulic conditions. To support the design of the flow test for MURR LEU fuel element hydraulic performance evaluation, design parameters for hydraulic testing of the LEU fuel element are laid out in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Data Report on the Newest Batch of PCEA Graphite for the VHTR Baseline Graphite Characterization Program

This report details a comparison of mechanical and physical properties from the first billet of extruded PCEA nuclear-grade graphite from the newest batch of PCEA procured from GrafTech. Testing has largely been completed on three of the billets from the original batch of PCEA, with data distributions for those billets exhibiting a much wider range of values when compared to the distributions of properties from other grades. A higher propensity for extremely low values or specimens that broke while machining or handling was also characteristic of the billets from the first batch, owing to unusually large fissures or disparate flaws in the billets in an as-manufactured state. Coordination with GrafTech prior to placing the order for a second batch of PCEA included discussions on these large disparate flaws and how to prevent them during the manufacturing process. This report provides a comparison of the observed data distributions from properties measured in the first billet from the new batch of PCEA with those observed in the original batch, in order that an evaluation of tighter control of the manufacturing process and the outcome of these controls on final properties can be ascertained. Additionally, this billet of PCEA is the first billet to formally include measurements from two alternate test techniques that will become part of the Baseline Graphite Characterization database – the three-point bend test on sub-sized cylinders and the Brazilian disc splitting tensile strength test. As the program moves forward, property distributions from these two tests will be based on specimen geometries that match specimen geometries being used in the irradiated Advanced Graphite Creep (AGC) program. This will allow a more thorough evaluation of both the utility of the test and expected variability in properties when using those approaches on the constrained geometries of specimens irradiated in the Advanced Test Reactor as part of the AGC experiment.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effect of alloying elements, cold work, and hydrogen on the irradiation-induced growth behavior of zirconium alloy variants - 2016-0040

In-reactor dimensional changes in zirconium-based alloys result from a complex interplay of many factors, such as (1) alloy type and composition, including the addition of elements such as niobium, iron, and tin; (2) fabrication process, including cold work, texture, and residual stresses; (3) irradiation temperature; and (4) hydrogen levels. In many cases, the observed dimensional changes in light water reactor fuel-assembly components-especially at high exposures-cannot be fully explained based on current growth and creep models. Therefore, a systematic approach was taken in this multi-year (2005-2011) Nuclear Fuel Industry Research Program investigation. The objective was to measure stress-free irradiation-induced growth (IIG) of specially fabricated alloys through irradiation under controlled conditions in the BOR-60 fast-flux test reactor up to a high fluence of approximately 2X10{sup 26} m{sup -2} (E>1 MeV)-equivalent to maximum of approximately 37 dpa exposure-followed by postirradiation examinations (PIEs). Irradiation temperature was within a narrow temperature range (320±10 deg. C). The PIEs included dimensional-change and microhardness measurements, metallography and hydride etching, and scanning transmission electron microscopy (STEM) or transmission electron microscopy (TEM). All irradiation samples (typically flat rectangular coupons or curvilinear cutouts of cladding tubes sized 35 by 6.5 by 0.8 mm) were pre-filmed to avoid the uptake of impurity hydrogen from sodium-cooled BOR-60. A wide variety of samples representing standard LWR cladding alloys with and without pre-hydriding (approximately 116 to approximately 718 ppm) as well as special compositions with iron contents (100 to 4,000 ppm) were irradiated. The irradiation in BOR-60 was done in five different stages (eight micro-cycles) and lasted approximately 18 months with interim and final growth measurements made using a high precision-length measurement device. Results of the extensive investigation include: significant effects of Fe, Nb, and hydrogen additions; quantification of growth rates from low to very high fluences (dpas); measurement of volume changes; and correlation of growth with -component dislocation densities. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2017

Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in BWR Conditions

As a first step toward full scale utilization of the Advanced Test Reactor (ATR) and associated post irradiation examination (PIE) equipment at INL, the NRC and INL staffs have formulated a test program that is designed to help to establish INL and the Nuclear Science User Facilities (NSUF) as a viable destination for irradiation and PIE of reactor structural materials. This initial research program utilized two materials (304 Stainless Steel weld Heat Affected Zone (HAZ) and sensitized 304L stainless steel) that have previously been irradiated at the Halden reactor in Norway and tested at Argonne National Laboratory (ANL) so that test results may be compared, thereby establishing a measure of comparability between Halden and ATR irradiations. A secondary objective of this test program was to characterize the quality of data produced using INL’s newly constructed irradiation assisted stress corrosion cracking (IASCC) test cells. The specimens were be tested in typical boiling water reactor (BWR) conditions to measure stress corrosion cracking (SCC) and fracture toughness of un-irradiated specimens and then IASCC and fracture toughness of specimens irradiated to a fluence equal to approximately 1.0 X 10 21 n/cm 2 (E > 1 MeV).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1972

Development of a model and test equipment for cold flow tests at 500 atm of small nuclear light bulb configurations

A model and test equipment were developed and cold-flow-tested at greater than 500 atm in preparation for future high-pressure rf plasma experiments and in-reactor tests with small nuclear light bulb configurations. With minor exceptions, the model chamber is similar in design and dimensions to a proposed in-reactor geometry for tests with fissioning uranium plasmas in the nuclear furnace. The model and the equipment were designed for use with the UARL 1.2-MW rf induction heater in tests with rf plasmas at pressures up to 500 atm. A series of cold-flow tests of the model was then conducted at pressures up to about 510 atm. At 504 atm, the flow rates of argon and cooling water were 3.35 liter/sec (STP) and 26 gal/min, respectively. It was demonstrated that the model is capable of being operated for extended periods at the 500-atm pressure level and is, therefore, ready for use in initial high-pressure rf plasma experiments.

Jaminet, J. F.↗
DOE OSTI2026

Beyond interpolation: Physics-inspired gating transformers for extrapolating irradiation conditions to novel nuclear fuels

The qualification of advanced nuclear fuels relies on irradiation experiments in test reactors that emulate commercial conditions. Designing these tests requires accurate prediction of key irradiation quantities, particularly heat generation rate and burnup, yet obtaining them typically involves computationally expensive multi-step simulation workflows. We propose a physics-inspired gating transformer (PIGT) that integrates an inverse-square, distance-based attenuation into the encoder representation to bias attention toward physically relevant spatial relationships while retaining data-driven flexibility. Using MiniFuel irradiation data from the High Flux Isotope Reactor at Oak Ridge National Laboratory, we benchmark against ensemble methods, feedforward and recurrent networks, convolutional models, and standard transformers. While baseline models perform well under interpolation, they exhibit a pronounced generalization gap when evaluated on fuels not included in the training set. The proposed model consistently improves extrapolative accuracy and stability, yielding the strongest performance on unseen fuel configurations. These results indicate that a lightweight physics structure embedded within attention mechanisms can substantially improve robustness, enabling more reliable surrogate predictions to accelerate the design of nuclear fuel irradiation experiments.

Fuel qualification↗
DOE OSTI2017

Upcoming irradiation of ultrasonic and fiber optic sensors in the MITR - 361

Fiber optic and ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. These technologies, which have performed successfully in out-of-pile tests, have not been well qualified in a test reactor environment. The uncertainty of successful operation of instruments at this level of development has typically been a significant barrier to in-core deployment. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of these technologies to provide enhanced sensors for irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. The National Scientific User Facility (NSUF) is sponsoring a project to evaluate the performance of promising temperature sensors based on ultrasound and on fiber optics during an upcoming irradiation test in the Massachusetts Institute of Technology Research Reactor (MITR). Several types of thin-wire metallic waveguide ultrasonic thermometers will be tested along with fiber optic temperature sensors using regenerated Fiber-Bragg Grating technology. Additionally, samples of several new optical materials which are potentially much more radiation tolerant than commonly used fiber materials, such as silica, will be included and analyzed post-irradiation. The irradiation is a collaborative effort between researchers from Idaho National Laboratory, the French Alternative Energies and Atomic Energy Commission, the University of Pittsburgh, the Massachusetts Institute of Technology, and AFO Research, Inc. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Effect of Water Chemistry on Crack Growth Rate in Neutron Irradiated X-750 and XM-19

The Advanced Test Reactor (ATR) National Science User Facility (NSUF) was used to study the effect of neutron damage on crack growth rates (CGR) under simulated boiling water reactor (BWR) water chemistry conditions using a fast neutron spectrum to accelerate the damage process. Nickel based Alloy X-750 and XM-19 were irradiated in the ATR center flux trap (CFT), with this experiment marking the first civilian project to utilize this position within the reactor. 0.4T-CT specimens were used to study crack growth rates for two sample fluences (~1.9 x 1020 n/cm2 and ~9.5 x 1020 n/cm2, E > 1MeV)) in both normal water chemistry (NWC) at 2.5 ppm O2 and hydrogen water chemistry (HWC) at ~80 ppb H2. CGR testing was conducted at a temperature of 288 °C. A decrease in CGR was observed for both alloys with the average ratio (CGR in NWC/CGR in HWC) being a factor of 281 for the XM-19 samples and a factor of 27.8 for the X-750 samples. Observed CGRs in this study showed little difference compared to non-irradiated samples (the non-irradiated XM-19 was 20% cold worked), suggesting limited sensitivity of these susceptible alloys to irradiation effects. Detailed results of CGR response to varying environmental electrochemical potential (ECP) and water purity will be presented in the context of crack growth mitigation in different water chemistries and the change in material behavior relative to non-irradiated samples of the material.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Poster for 12th World Conference for Neutron Radiography (WCNR-12)

Complimentary Relationships Between Neutron Computed Radiography and Digital X-ray Radiography of Radioactive Specimens The Transient Reactor Test (TREAT) Facility is a 19,000 MW reactor located at the Idaho National Laboratory which is used for transient testing of materials and experiments. A neutron radiography facility is installed which allows for low resolution neutron imaging with a capability of items up to 20 cm wide and 400 cm long. Along with the neutron radiography facility, a 370 kV digital X-ray system is located onsite for when X-ray imaging is more practical or when rapid imaging is requested. The X-ray facility can accommodate specimens up to 61 cm wide and 1000 cm long. This poster discusses the complimentary relationship between transfer-method Neutron Computed Radiography (nCR) and Digital X-ray Radiography (DR) imaging of radioactive specimens along with practical applications for nuclear fuels and materials testing programs. Both nCR and DR images were acquired of inert gas and water filled apparatus containing differing materials at varying radiation levels. Highly radioactive items have no effect on nCR image quality whereas radioactivity has varying degrees of effect in DR. Radioactive items as high as 20 mSv/h have been successfully imaged using X-ray DR methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Accident tolerant, non-oxidizable MOX fuel fabrication

Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Three-dimensional FIB/EBSD characterization of irradiated HfAl{sub 3}-Al composite

A thermal neutron-absorbing composite comprised of 28.4 vol.% HfAl{sub 3} intermetallic particles in an Al matrix, developed for use in a conduction-cooled thermal neutron filter, will enable fast flux irradiation of materials and fuels in an existing pressurized water test reactor. In this paper, an EBSD-FIB characterization approach is used to observe the microstructural changes of the HfAl{sub 3}-Al composite in response to simulated long-term neutron irradiation. Using a focused ion beam (FIB), the sample was fabricated to 25 μm x 25 μm x 20 μm and mounted on a transmission electron microscopy (TEM) grid. A series of operations were carried out repeatedly on the sample's top surface to prepare it for scanning electron microscopy (SEM). First, a ∼ 100-nm layer was removed by high voltage FIB milling. Then, several cleaning passes were performed on the newly exposed surface using low voltage FIB milling to improve the SEM image quality. The surface was then scanned by Electron Backscattering Diffraction (EBSD) to obtain a two-dimensional image. After 50 to 100 two-dimensional images were collected, the images were stacked to reconstruct a three-dimensional model using DREAM.3D software. Two such reconstructed three dimensional models were obtained from samples of the original and post-irradiation HfAl{sub 3}-Al composite, from which the most significant microstructural change caused by neutron irradiation appears to be the size reduction of both HfAl{sub 3} and Al grains. Possible causes include thermal expansion and related thermal strain from thermal neutron capture. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

TCR Central Shutdown Rod Fine Motion Control

Transformational Challenge Reactor (TCR) is a Helium cooled 3 MWt test reactor that leverages advances in materials and manufacturing, computing, and AI in its design. Classical design of a shutdown rod uses either gravity, pneumatic, or springs to quickly release the Central Shutdown Rod (CSR) containing neutron absorber into the reactor core stopping nuclear reaction. Generally, the motor/actuator resides outside the reactor, but the motion is transmitted through a penetration into the pressure vessel. There are also designs where the control rod drive incorporates magnetic latches with coil residing outside the pressure boundary for precise position control of the rod. We are proposing a magnetic coupling to position the shutdown rod without any penetration into the pressure vessel for the entire drive length of TCR shutdown rod. Electromagnets are currently used in non-power nuclear reactors for shutdown rods, but these electromagnets are resident inside the reactor pressure vessel. This paper will describe the use of an electromagnet outside the pressure vessel to position and release the shutdown rod. A prototype was developed at ORNL to demonstrate the concept, and a design optimization of the ferritic core and material was conducted to maximize the lift force of the electromagnet. An elevated temperature testing was also performed to ensure that the system will perform under the temperature conditions inside an operating reactor.

Fountain, Eliott J.↗
DOE OSTI2019

Hybrid super homogenization and discontinuity factor method for continuous finite element diffusion

This paper presents a novel homogenization equivalence technique aiming to simultaneously leveragethe simplicity of the Super Homogenization (SPH) method to reproduce reference reaction rates andthe ability to preserve reference leakage rates at desired surfaces through Discontinuity Factors (DF).The need for this new class of methods arises from the inability of the current state-of-the-art SPH technology to properly reproduce the reactor multiplication factor for problems with significant leakage. This work shows that this defect lies in the use of normalization factors in the SPH algorithm: while they are necessary with purely reflecting problems to ensure uniqueness of the solution, they introduce homogenization inconsistencies if at least one vacuum boundary condition is present. Two solutions to this problem are presented in this work: (i) to simply remove the normalization factor or (ii) to introduce additional degrees of freedom in the form of DFs in such a way that the normalization factors can still constrain the problem. While the former clearly offers unrivaled simplicity and can lead to very satisfactory results for leakage-dominated cores or if spatial restriction of the SPH regions is applied – as demonstrated for the High Temperature Test Reactor, the latter is more robust and does not require to scale the fission terms with the reference eigenvalue.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Enhancements to the New TREAT Automatic Reactor Control System (ARCS)

The Transient Reactor Test (TREAT) facility recently replaced the automatic reactor control system (ARCS). Here, this paper discusses some of the enhancements that were made during the ARCS replacement such as calculations to take into account the nonadiabatic effects and increasing the dynamic range for power and period indication. This paper also discusses upgrades that had been planned but were unable to implemented and potential upgrades for the future.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Assessment of Pump Coastdown and Reactor Protection Scram Delays during Station Blackout Conditions

This report presents an assessment of the impact of pump coastdown time and reactor protection scram delays for the Versatile Test Reactor (VTR) during Protected Station Blackout (PSBO) conditions. While previous simulations for the VTR assumed centrifugal pumps in the primary system, the current pre-conceptual design uses electromagnetic (EM) pumps. Because there is uncertainty as to what the parameters of the EM pump design will be, this report focuses on the impact of the flow coastdown and the reactor protection scram delay, not on the needs of the pump design. PSBO transient simulations were performed using the SAS4A/SASSYS-1 fast reactor safety analysis code (version 5.2.3). The model used in the simulations is based on the FY18 SAS4A/SASSYS-1 model. A change to the model was made to replace the centrifugal pump model with an electromagnetic pump model with a motor generator as a coastdown mechanism. The results of the PSBO transient simulations were assessed against the criteria used in Reference 2 with an additional 50°C for conservatism to account for uncertainties. The results show that as the delay before scramming the rods increases, the moment of inertia of the coastdown mechanism must increase to satisfy the criteria. By increasing the moment of inertia to extend the flow coastdown, temperatures can be kept below the desired limits with reactor protection scram delays between 3 and 10 seconds.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Characterization of Aluminum Metal Powders for Plutonium-238 Program

As part of radioisotope thermoelectric generator (RTG) production for power sources in deep space exploration, Oak Ridge National Laboratory (ORNL) produces plutonium-238 ( 238 Pu). To produce 238 Pu, neptunium-237 ( 237 Np) targets are fabricated at ORNL and subsequently irradiated at the High Flux Isotope Reactor (HFIR) at ORNL and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Specifically, current flowsheets utilize neptunium dioxide ( 237 NpO 2 ) targets. Aluminum (Al) powder is blended with 237 NpO 2 target material to improve thermal conductivity during irradiation. The current composition of 237 NpO 2 /Al pellets (i.e., cermets, or ceramic–metallics) is 20% 237 NpO 2 , 70% Al, and 10% void space. Al powders utilized by the 238 Pu program are high-fired under vacuum before blending into targets. The primary objective of this work is to measure the physical, chemical, and thermal properties of Al before and after the high-firing process. Properties of interest include the morphology, surface texture, particle size, crystal phase(s), and thermal conductivity of the material. These properties are measured with existing materials characterization equipment at ORNL, including powder x-ray diffraction (pXRD), scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS), laser flash analysis (LFA), and thermogravimetric analysis (TGA). A secondary objective of this work is to compare the Al powders before and after the high-firing procedure to determine the effects of high-firing on chemical, physical, and thermal properties and to determine the efficacy of the high-firing process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Metallurgical Analysis of the High Flux Isotope Reactor (HFIR) Carrier Lifting Bails (Rev.1)

The dissolution rates of the aluminum alloys in the High Flux Isotope Reactor (HFIR) element carriers and the Material Test Reactor (MTR) L-bundles in the H-Canyon facility have been identified as the possible cause of extended dissolutions that result in significant time and financial expenditures. A study, carried out by Savannah River National Laboratory (SRNL) to determine relationships between the dissolution rates and the metallurgical properties of the aluminum alloy materials of construction of the HFIR carriers and the L-bundles, considered the dissolution rates of aluminum alloy (AA) series 1100, 6061, and 6063. The study determined that the aluminum alloy compositions played a principal role in the dissolution rate of the carrier/bundle components. Higher dissolution rates were correlated with lower concentrations of the minor element additions in the alloys and with specific element concentrations. Aluminum alloys 1100 and 6063 were found to have similar dissolution rates that were approximately two orders of magnitude (100X) greater than those of AA6061. Based on the results of the dissolution behavior study, a Technical Assistance Request (TAR) was first issued to determine if the replacement of AA6061-T6 with AA6063-T6 is feasible for the HFIR carrier lifting bails. A Technical Task Request was then issued to consider AA6063-T5 as well as other alloys to improve possible supply chain issues. The metallurgical properties of the L-bundle (specifically the end caps) were not evaluated in this report because L-Bundle drawings already allow for the use of AA6063-T6 in all structural components. The HFIR carriers are composed of thin-walled components with significant surface areas that allow for relatively quick overall dissolution times. Conversely, the carrier lifting bails and the supporting constituents are composed of solid bars and thick plate regions with relatively small surface areas that experience longer overall dissolution times. While the MTR L-bundle design includes allowances for the materials of construction to be either AA6061-T6 or AA6063-T6, the HFIR carriers are specified to be constructed fully with AA6061-T6 alloy. This report analyzes the recommendations of the dissolution behavior study to replace the materials of construction of the HFIR carrier lifting bails. The analysis considers the operational requirements of the lifting bail and its supporting structures. To decrease dissolution times, the analysis considers direct replacement of the material as well as reductions in the thicknesses of the components to decrease the mass of the elements. Material reductions are considered on options for using either AA6061 and/or AA6063. The calculations are based on specifications from the American Society of Mechanical Engineer (ASME) and The Aluminum Association, Inc. design codes. The analysis finds that direct replacement of the lifting bail material of construction with AA6063-T6, and AA6063-T5 as well as reductions in the dimensions of the lifting bail components are acceptable. Note that this study considers the structural suitability of the alloys. It does not consider their dissolution rates in the dissolvers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗