The Combinatorial Approach to Testing and Characterization of Irradiated Fuels and Reactor Structural Materials
Testing and characterization of irradiated fuels and reactor structural materials
36 - MATERIALS SCIENCE↗
SEARCH · Search NASA
Results for “REACTOR MATERIALS”
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.
Dissolution of Material Test Reactor Fuel in an H-Canyon Dissolver
In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO 3 flowsheet. Since the development of the existing flow sheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report
This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
A Plan to Qualify New Fuel for the High Flux Isotope Reactor for Material Minimization Poster
A Plan to Qualify New Fuel for the High Flux Isotope Reactor for Material Minimization Poster for presenting at Waste Management Symposium 2023, which starts 2/26/2023.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thick Films acoustic sensors devoted to MTR environment measurements. Thick Films acoustic sensors devoted to Material Testing Reactor environment measurements
The development of advanced instrumentation for in-pile experiments in Material Testing Reactor constitutes a main goal for the improvement of the nuclear fuel behavior knowledge. An acoustic method for fission gas release detection was tested with success during a first experiment called REMORA 3 in 2010 and 2011, and the results were used to differentiate helium and fission gas release kinetics under transient operating conditions. This experiment was lead at OSIRIS reactor (CEA Saclay, France). The maximal temperature on the sensor during the irradiation was about 150 deg. C. In this paper we present a thick film transducer produce by screen printing process. The screen printing of piezoelectric offers a wide range of possible applications for the development of acoustic sensors and piezoelectric structure for measurements in high temperature environment. We firstly produced a Lead Zirconate Titanate (PZT) based paste composed of Pz27 powder from Ferroperm, CF7575 glass, and organic solvent ESL 400. Likewise a Bismuth Titanate based paste synthesized in our laboratory was produced. With these inks we produced thick film up to 130 μm by screen printing process. Material properties characterizations of these thick-film resonators are essential for device design and applications. The piezoelectric coefficients d33 and pyro-electric P(T) coefficient are investigated. The highest P(T) and d33 are respectively 80 μC.m{sup -2}.K{sup -1} and 130 μC.N{sup -1} for the PZT transducer -which validates the fabrication process-. In view of the development of this transducer oriented for high temperature and irradiation environment, we investigated the electrical properties of the transducers for different ranges of frequencies and temperature - from 20 Hz up to 40 MHz between 30 and 400 deg. C. We highlight the evolution of the impedance response and piezoelectric parameters of screen printed piezoelectric structures on alumina. Shortly an irradiation will be realized in order to investigate their reliability towards neutron fluence. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors
An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Particle suspension reactors and materials for solar-driven water splitting
Reactors based on particle suspensions for the capture, conversion, storage, and use of solar energy as H 2 are projected to be cost-competitive with fossil fuels. In light of this, this review paper summarizes state-of-the-art particle light absorbers and cocatalysts as suspensions (photocatalysts) that demonstrate visible-light-driven water splitting on the laboratory scale. Also presented are reactor descriptions, theoretical considerations particular to particle suspension reactors, and efficiency and performance characterization metrics. Opportunities for targeted research, analysis, and development of reactor designs are highlighted.
14 SOLAR ENERGY↗
Processing of Nanostructured Ferritic Alloys for Advanced Reactor Core Materials
The Fuel Cycle Research and Development Program is investigating methods of burning minor actinides in a transmutation fuel. To achieve this goal, the fast reactor core materials (cladding and duct) must be able to withstand very high doses (>200 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI). One of the most promising materials for this demanding applications are the class of nanostructured ferritic alloys. These materials are ferritic alloys with high chrome (9-14Cr) with a high density (>10{sup 23}/m{sup 3}), fine distribution of nanosized (∼2 nm) oxide particles and a fine grain size (<0.5 micron). This fine microstructure provides an alloy with high strength at high temperatures and excellent radiation tolerance (e.g. reduced void swelling and ductility retention at low temperatures) but also increases the difficulty of producing engineering parts (e.g. thin walled tubes) from these advanced materials. Thus, extensive research is underway to produce tubes using techniques such as high temperature hydrostatic extrusion, intermediate temperature plug drawing and low temperature pilger processing. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Overview/Recap of the 2020 Accelerated Irradiation of Reactor Structural Materials Workshop
The workshop provided a snapshot look at current international, in-reactor testing and irradiation capabilities. Gaps were identified where possible, and discussions focused on potential mitigation strategies and recommended paths forward. As new material innovations are being developed for in-reactor applications for reactor life extension long-term operation, and advanced reactor technologies, there is an increased need for materials qualification and assessment programs. In-reactor testing capabilities are vital to the on-going success of these DOE-NE programs and initiatives. In recent years, multiple international research reactors have been placed in a permanent shut-down state and have begun full decommissioning. These activities have created new risks in the delivery of on-going and/or expected research programs. Beyond performing simple irradiations in test reactors, few facilities exist internationally which can perform instrumented, in-situ irradiations on structural materials, and with the recent shut down of facilities hosted at Halden and NRU (example: instrumented fatigue loop, and in-situ creep), there are further gaps in the industry left un-filled. A new focus is being placed on the use of accelerator- based technologies to fill in some of these gaps, but these must be viewed as supplemental, and not surrogates to in-reactor capabilities
36 - MATERIALS SCIENCE↗
An overview of the 2014 revision of ANSI/ANS-ANS-8.1-2014, 'Nuclear criticality safety in operations with Fissionable Materials Outside Reactors' - 14577
A revision to the American National Standards Institute (ANSI)/American Nuclear Society (ANS) standard, 'Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors' [ANSI/ANS-8.1 (ANS-8.1)], was completed and published in 2014. The approved scope for the 2014 revision included two major efforts: (1) an editorial revision to clarify the meaning of the process analysis requirement (PA) and the double contingency principle recommendation (DCP) and (2) the generation of new subcritical limits for uranium and uranium compounds for enrichments less than 10 wt % {sup 235}U. After a clarification request was submitted regarding the PA and DCP in 2009, the scope of the revision became focused almost entirely on the editorial revision. With approval from the ANS-8 subcommittee chairman, the computational effort was postponed until the editorial revision could be completed. The 2014 revision of ANS-8.1 involved nuclear criticality safety experts from 14 different sites, each with many years of experience, either maintaining the standard as a working group member or as a criticality safety practitioner. Many of the represented sites had implemented the PA with the language of the DCP to ensure that process operations with fissionable materials would be subcritical. However, the early working group deliberations made it clear that the general industry focus was on the DCP rather than the PA, which requires that 'all normal and credible abnormal conditions remain subcritical'. Over the course of the revision, two new definitions and a new appendix (Appendix B) were developed to enhance understanding of the PA and DCP. The emphasis of Appendix B is on the requirement for subcriticality (PA) rather than the DCP. The phrase 'credible abnormal conditions' in the PA is broader in meaning than 'changes in process conditions,' as recommended by the DCP. This is because a single credible abnormal condition can potentially affect multiple independent process conditions, and consequently, multiple process parameters, simultaneously. The DCP is a technical practice meant to support the PA and cannot be applied to all situations. However, when it can be properly applied, the DCP technical practice, in most cases, ensures compliance with the PA because two unlikely, independent, and concurrent changes in process conditions are generally judged not to be a credible abnormal condition, as specified in the PA. This paper elaborates on the intense working group discussions over the course of the ANSI/ANS-8.1 standard revision and summarizes the modifications made to the revised standard. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Accident Performance of Light Water Reactor Cladding Materials
During a loss of coolant accident as experienced at Fukushima, inadequate cooling of the reactor core forces component temperatures ever higher where they must withstand aggressive chemical environments. Conventional zirconium cladding alloys will readily oxidize in the presence of water vapor at elevated temperatures, rapidly degrading and likely failing. A cladding breach removes the critical barrier between actinides and fission products and the coolant, greatly increasing the probability of the release of radioactivity in the event of a containment failure. These factors have driven renewed international interest in both study and improvement of the materials used in commercial light water reactors. Characterization of a candidate cladding alloy or oxidation mitigation technique requires understanding of both the oxidation kinetics and hydrogen production as a function of temperature and atmosphere conditions. Researchers in the MST division supported by the DOE-NE Fuel Cycle Research and Development program are working to evaluate and quantify these parameters across a wide range of proposed cladding materials. The primary instrument employed is a simultaneous thermal analyzer (STA) equipped with a specialized water vapor furnace capable of maintaining temperatures above 1200°C in a range of atmospheres and water vapor contents. The STA utilizes thermogravimetric analysis and a coupled mass spectrometer to measure in situ oxidation and hydrogen production of candidate materials. This capability is unprecedented in study of materials under consideration for reactor cladding use, and is currently being expanded to investigate proposed coating techniques as well as the effect of coating defects on corrosion resistance.
08 HYDROGEN↗
JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity
Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and D support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under construction at CEA Cadarache research centre in the south of France will represent a major Research Infrastructure for scientific studies regarding material and fuel behavior under irradiation. It will also be devoted to medical isotopes production. Hence JHR will offer a real opportunity to perform R and D programs regarding needs above and hence will crucially contribute to the selection, optimization and qualification of these innovative materials and fuels. The JHR reactor objectives, principles and main characteristics associated to specific experimental devices associated to measurement techniques and methodology, their performances, their limitations and field of applications will be presented and discussed. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Fusion-Driven Subcritical System FDS-SFB and Its Blanket Materials Activation Analysis
Fusion-fission hybrid reactor is a bridge between fusion and fission nuclear energy and it is a solution to overcome difficulties of fission energy development with the features of energy production, fuel breeding and waste transmutation. Many researches on fusion-fission hybrid reactor have been done in last decades. Fusion Driven Subcritical Spent Fuel Burning System (FDS-SFB) was proposed as an intermediate step toward the final application of fusion energy and designed to transmute the long-lived nuclear wastes from fission power plants and to produce fissile nuclear fuels for feeding fission power plants. The design of FDS-SFB is based on relatively easy-achieved plasma parameters extrapolated form the successful operation of the experimental advanced superconducting Tokamak (EAST) in China and other tokamaks in the world, and the subcritical blanket is designed based on the well-developed technologies of fission reactor. Materials activation characteristic is one of the key issues for hybrid reactor research. The activation of materials would substantially influence not only the safety of the reactors but also the recycling of used reactor materials. The selection of materials is very important to control the neutron-induced activation on fusion reactor as well as to ensure the attractiveness of fusion-fission nuclear power regarding safety and environmental aspects. The reduced activation ferritic-martensitic (CLAM) steel is preliminarily considered as the candidate structural materials for FDS-SFB. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Solubility Testing to Support the Addition of Sodium Reactor Experiment Material to Sludge Batch 10
The Savannah River Site H-Canyon Facility is planning to discard dissolved Sodium Reactor Experiment (SRE) material into Tank 51 prior to Sludge Batch 10 (SB10). SB10 with the SRE material will be processed in the Defense Waste Processing Facility (DWPF) using the Nitric-Glycolic Acid (NGA) flowsheet. The DWPF Nuclear Criticality Safety Evaluation (NCSE) protects a 14:1 mass ratio of manganese to equivalent uranium-235 in both the solid and liquid phases during processing by requiring a 70:1 mass ratio in the feed. A concern was raised that freshly precipitated manganese in the SRE material will behave differently from the manganese in sludge during DWPF processing and potentially invalidate the criticality control. To mitigate risks to the DWPF criticality controls, the Savannah River National Laboratory (SRNL) performed a metals solubility test applicable to recently precipitated manganese and uranium, as would be expected with SRE transfers to SB10. The objective of this analysis is the tracking of partitioning of the primary fissile component (uranium-235) with the credited poison (manganese) between the aqueous and insoluble phases.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Corrosion Testing Needs and Considerations for Additively Manufactured Materials in Nuclear Reactors
The Advanced Materials and Manufacturing Technologies (AMMT) program within the Department of Energy, Office of Nuclear Energy (DOE-NE) has developed its current recommendation for its corrosion testing strategy to deploy additively manufactured (AM) materials in advanced nuclear reactors. Additive manufacturing technologies have developed rapidly in recent years, creating new opportunities and challenges for the nuclear industry. To adopt AM technologies, the corrosion performance of AM materials needs to be adequately evaluated.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Effect of surface oxidation on the onset of nucleate boiling in a materials test reactor coolant channel
The onset of nucleate boiling (ONB) serves as the thermal-hydraulic operating limit for many research and test reactors. However, boiling incipience under forced convection has not been well-characterized in narrow channel geometries or for oxidized surface conditions. This study presents experimental data for the ONB in vertical upflow of deionized (DI) water in a simulated materials test reactor (MTR) coolant channel. The channel gap thickness and aspect ratio were 1.96 mm and 29:1, respectively. Boiling surface conditions were carefully controlled and characterized, with both heavily oxidized and native oxide surfaces tested. Measurements were performed for mass fluxes ranging from 750 to 3000 kg/m 2 s and for subcoolings ranging from 10 to 45°C. ONB was identified using a combination of high-speed visual observation, surface temperature measurements, and channel pressure drop measurements. Surface temperature measurements were found to be most reliable in identifying the ONB. For the nominal (native oxide) surface, results indicate that the correlation of Bergles and Rohsenow, when paired with the appropriate single-phase heat transfer correlation, adequately predicts the ONB heat flux. Furthermore, incipience on the oxidized surface occurred at a higher heat flux and superheat than on the plain surface.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Plan for Post-Irradiation Examination of Zion Reactor Pressure Vessel Beltline Materials
The reactor pressure vessel (RPV) is a potentially life-limiting component in light-water reactors (LWR) because replacement of the RPV is not considered a viable option. Researchers studying the effects of radiation on RPV materials have long been interested in evaluating service-irradiated materials to assess currently available models for predicting radiation embrittlement of RPV steels. The decommissioning of the Zion Unit 1 NPP in Zion, Illinois presents a unique opportunity for characterization of an actual reactor pressure vessel (RPV) material after inservice degradation. The main interests in this case are the belt-line weld and base metals. Through-RPV wall thickness attenuation and chemical composition distributions are the obvious objectives of this characterization program. However, it also provides an opportunity to examine the inservice degradation and its comparison with currently available models for prediction of radiation embrittlement of RPV steels. All of these efforts will provide a better understanding of materials degradation and other issues associated with extending the lifetime of existing NPPs beyond 60 years of service. The primary foci are the Linde 80 flux, wire heat 72105 (WF-70) circumferential belt-line weld and the A533B-1 base metal from the intermediate shell harvested from a region of peak fluence (0.7 x 10{sup 19} n/cm{sup 2}, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing will be being performed to characterize the through-thickness mechanical properties of the base metal and belt-line-weld materials. In addition to mechanical properties, microstructural characterizations will be performed using various techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗