Effects of Thermal Treatment on the Co-rolled U-Mo Fuel Foils

ABSTRACTA monolithic fuel design based on U–Mo alloy has been selected as the fuel type for conversion of United States’ high-performance research reactors (USHPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184kJ/mole, consistent with a previous diffusion-couple study. The phases in the U–Mo/Zr interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion-couple study.

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Innovative and safe supply of fuels for reactors

EPJ Nuclear Sciences & Technologies ◽

10.1051/epjn/2019013 ◽

2020 ◽

Vol 6 ◽

pp. 40

Author(s):

Stéphane Valance ◽

Bruno Baumeister ◽

Winfried Petry ◽

Jan Höglund

Keyword(s):

Power Plant ◽

Nuclear Power ◽

High Performance ◽

Research Reactor ◽

Fissile Material ◽

Nuclear Fuels ◽

Research Reactors ◽

Enriched Uranium ◽

Performance Research ◽

And Training

Within the Euratom research and training program 2014–2018, three projects aiming at securing the fuel supply for European power and research reactors have been funded. Those three projects address the potential weaknesses – supplier diversity, provision of enriched fissile material – associated with the furbishing of nuclear fuels. First, the ESSANUF project, now terminated, resulted in the design and licensing of a fuel element for VVER-440 nuclear power plant manufactured by Westinghouse. The HERACLES-CP project aimed at preparing the conversion of high performance research reactor to low enriched uranium fuels by exploring fuels based on uranium-molybdenium. Finally, the LEU-FOREvER pursues the work initiated in HERACLES-CP, completing it by an exploration of the high-density silicide fuels, and including the diversification of fuel supplier for soviet designed European medium power research reactor. This paper describes the projects goals, structure and their achievements.

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A Neutronics Feasibility Study of the TRIGA LEU Fuel in the 20MWt NIST Research Reactor

Volume 9: Student Paper Competition ◽

10.1115/icone26-82433 ◽

2018 ◽

Cited By ~ 1

Author(s):

Kyle Anthony Britton ◽

Zeyun Wu

Keyword(s):

Feasibility Study ◽

High Performance ◽

National Bureau ◽

Fuel Type ◽

Research Procedure ◽

Research Findings ◽

Enriched Uranium ◽

Global Threat ◽

Performance Research ◽

Neutronic Performance

The National Bureau of Standards reactor (NBSR) at the National Institute of Standards and Technology (NIST) is under conversion from high enriched uranium (HEU) to the low enriched uranium (LEU) schema under the Reduced Enrichment for Research and Test Reactors program (RERTR) as a part of the Global Threat Reduction Initiative (GTRI). The conversion of the high performance research reactors (HPRR) such as NBSR is a challenging task due to the high flux need (2.5 × 1014 n/cm2-s for the NBSR), as well as other neutronics performance characteristics requirements without significant changes to the external geometrical configuration. One fuel candidate, the General Atomics (GA) UZrH LEU fuel, has showed particular promise in this regard. The TRIGA LEU fuel was initially developed in the 1980s with particular considerations for fuel conversion for high power regimes such as high density research and test reactors. This study performs a neutronics feasibility study of the UZrH LEU fuel schema for the NBSR, examining the accountability and sustainability of the TRIGA fuel when applying it to the NBSR conversion. To identify the best option to deploy the TRIGA fuel to NBSR in terms of key neutronic performance characteristic, the study is carried out with various considerations in the fuel dimensions, fuel rod layout configurations, and structure material selections. Monte Carlo based computational model is used to assist and facilitate the research procedure. The research findings in this study will determine the viability of the TRIGA fuel type for the NBSR conversion, and provide supporting data for future investigations on this subject.

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Experiment and Design Integration for U.S. High Performance Research Reactor Low-Enriched Uranium Conversion

10.13182/t123-33150 ◽

2020 ◽

Author(s):

D. Jaluvka ◽

E. Wilson ◽

L. Jamison ◽

J. Stillman

Keyword(s):

High Performance ◽

Research Reactor ◽

Design Integration ◽

Enriched Uranium ◽

Performance Research

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Effects of Fission Profiles on the Performance of a U-10Mo Fuel Plate

ASME 2019 Power Conference ◽

10.1115/power2019-1895 ◽

2019 ◽

Author(s):

Hee Seok Roh ◽

Walid Mohamed ◽

Hakan Ozaltun

Keyword(s):

High Performance ◽

Mechanical Performance ◽

Experimental Tests ◽

Operating Conditions ◽

Fuel Performance ◽

Element Analysis ◽

Actual Use ◽

Enriched Uranium ◽

Performance Research ◽

Plate Geometry

Abstract In order to convert the high-performance research reactors from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel, U-Mo alloy-based fuels in monolithic form have been proposed. These plate-type fuels consist of a high density and low enriched uranium (LEU) foil coated with a diffusion barrier and encapsulated with the aluminum cladding. The performance of the fuel plate has been evaluated by many studies through experimental tests and numerical analyses. When evaluating the performance of a fuel, it is expensive and time-consuming to consider a variation of several parameters, such as fuel plate geometry, material properties, and operating conditions. Fission profile is a critical component of the fuel performance analysis, causing swelling and creep deformation of the fuel plate. Therefore, it can directly affect the stress and strain distributions over the fuel plate. This study aims at investigating the effect of different fission profiles on the thermo-mechanical performance of the fuel plate by finite element analysis. To investigate the effect of fission profile on fuel performance, several different fission profiles were generated and analyzed. The fission profiles were generated based on actual use.

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Status of the Low Enriched Uranium Fuel Development for High Performance Research Reactors

Advances in Science and Technology ◽

10.4028/www.scientific.net/ast.94.43 ◽

2014 ◽

Vol 94 ◽

pp. 43-54 ◽

Cited By ~ 1

Author(s):

Leo Sannen ◽

Sven van den Berghe ◽

Ann Leenaers

Keyword(s):

High Performance ◽

Alternative Fuels ◽

Volume Density ◽

Lessons Learned ◽

Research Reactors ◽

Dispersion Fuel ◽

Uranium Fuel ◽

Low Volume ◽

Enriched Uranium ◽

Performance Research

Historically, uranium enriched to >90% 235U has been used for many peaceful applications requiring high fission densities such as driver fuels for research reactors. However, the use of high-enriched uranium or HEU (all enrichments >20% 235U are considered HEU) for civil applications, is considered a proliferation concern. Since the 1970's, efforts are being devoted to the conversion of research reactors operating on HEU to alternative fuels using uranium with enrichment below 20% or LEU. These efforts imply the development of high-density LEU fuels to replace the low volume-density (mostly) UAlx based HEU fuels. The paper updates the present status of these developments focusing on the UMo dispersion fuel. It aims to provide an overview of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through irradiation experiments and post-irradiation examinations (PIE).

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Flow Testing and Analysis of the FSP-1 Experiment

ASME 2017 Nuclear Forum ◽

10.1115/nuclrf2017-3639 ◽

2017 ◽

Author(s):

Grant L. Hawkes ◽

Warren F. Jones ◽

Wade Marcum ◽

Aaron Weiss ◽

Trevor Howard

Keyword(s):

Flow Rate ◽

High Performance ◽

National Laboratory ◽

Water Channel ◽

Flow Characteristics ◽

Separate Branch ◽

Flow Test ◽

Enriched Uranium ◽

Test Reactor ◽

Fuel Meat

The U.S. High Performance Research Reactor conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Size Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water channel velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A pressure differential versus flow rate curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported.

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Development of High-Performance Copper Alloy Wire on a Continuous Casting and Rolling Process

Materia Japan ◽

10.2320/materia.55.114 ◽

2016 ◽

Vol 55 (3) ◽

pp. 114-116

Author(s):

Hitoshi Nakamoto ◽

Norikazu Ishida ◽

Yoshiaki Hattori ◽

Keiichiro Oishi ◽

Tetsuya Ashida

Keyword(s):

Continuous Casting ◽

Copper Alloy ◽

High Performance ◽

Rolling Process ◽

Alloy Wire

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ACR-1000™ Fuel Design Verification

18th International Conference on Nuclear Engineering: Volume 6 ◽

10.1115/icone18-30338 ◽

2010 ◽

Author(s):

David J. Wren ◽

Patrick Reid ◽

Len L. Wright

Keyword(s):

High Performance ◽

Reactor Core ◽

Cold Test ◽

Performance Criteria ◽

Extensive Experience ◽

Fuel Bundle ◽

Test Loop ◽

Fuel Design ◽

And Performance ◽

The Impact

The ACR-1000™ design is an evolutionary advancement of the proven CANDU® reactor design that delivers enhanced economic performance, safety, operability and maintainability. The fuel for the ACR-1000 design is based on the well established CANDU fuel bundle design that has over 40 years of demonstrated high performance. Building on its extensive experience in fuel design and analysis, and fuel testing, AECL has designed a CANFLEX-ACR™ fuel bundle that incorporates the latest improvements in CANDU fuel bundle design. The ACR-1000 fuel bundle also includes features that enable the ACR-1000 to achieve higher fuel burn-up and improved reactor core physics characteristics. To verify that the CANFLEX-ACR fuel bundle design will meet and exceed all design requirements, an extensive program of design analysis and testing is being carried out. This program rigorously evaluates the ability of the fuel design to meet all design and performance criteria and particularly those related to fuel failure limits. The design analyses address all of the phenomena that affect the fuel during its residence in the reactor core. Analysis is performed using a suite of computer codes that are used to evaluate the temperatures, deformations, stresses and strains experienced by the fuel bundle during its residence in the reactor core. These analyses take into account the impact of fuel power history and core residence time. Complementing the analyses, testing is performed to demonstrate the compatibility of the fuel with the reactor heat transport system and fuel handling systems, and to demonstrate the ability of the fuel to withstand the mechanical forces that it will experience during its residence in the core. The testing program includes direct measurement of prototype fuel element and fuel bundle properties and performance limits. A number of different test facilities are used including a cold test loop and a hot test loop with a full-scale ACR-1000 fuel channel that operates at reactor coolant temperatures, pressures and flows. This paper summarizes the out-reactor test program and related analysis that provide the basis for verifying that the ACR-1000 fuel design meets its requirements.

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Diffusion of Silicon Interstitials in thin Silicon Films on Insulator in Neutral and Oxidising Annealing Ambients

MRS Proceedings ◽

10.1557/proc-469-125 ◽

1997 ◽

Vol 469 ◽

Author(s):

Guénolé C.M. Silvestre

Keyword(s):

Integrated Circuits ◽

High Performance ◽

Annealing Treatment ◽

Silicon On Insulator ◽

Buried Oxide ◽

Thermal Oxide ◽

Good Crystalline Quality ◽

Thin Silicon ◽

Experimental Values ◽

Fault Growth

ABSTRACTSilicon-On-Insulator (SOI) materials have emerged as a very promising technology for the fabrication of high performance integrated circuits since they offer significant improvement to device performance. Thin silicon layers of good crystalline quality are now widely available on buried oxide layers of various thicknesses with good insulating properties. However, the SOI structure is quite different from that of bulk silicon. This paper will discuss a study of point-defect diffusion and recombination in thin silicon layers during high temperature annealing treatment through the investigation of stacking-fault growth kinetics. The use of capping layers such as nitride, thin thermal oxide and thick deposited oxide outlines the diffusion mechanisms of interstitials in the SOI structure. It also shows that the buried oxide layer is a very good barrier to the diffusion of point defects and that excess silicon interstitials may be reincorporated at the top interface with the thermal oxide through the formation of SiO species. Finally, from the experimental values of the activation energies for the growth and the shrinkage of stacking-faults, the energy of interstitial creation is evaluated to be 2.6 eV, the energy for interstitial migration to be 1.8 eV and the energy of interstitial generation during oxidation to be 0.2 eV.

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