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

2020 ◽  
Author(s):  
D. Jaluvka ◽  
E. Wilson ◽  
L. Jamison ◽  
J. Stillman
Keyword(s):  
High Performance ◽  
Research Reactor ◽  
Design Integration ◽  
Enriched Uranium ◽  
Performance Research

scholarly journals Innovative and safe supply of fuels for reactors

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.

Aqueous processing of U–10Mo scrap for high performance research reactor fuel

2012 ◽  
Vol 427 (1-3) ◽  
pp. 185-192
Author(s):  
Amanda J. Youker ◽  
Dominique C. Stepinski ◽  
Laura E. Maggos ◽  
Allen J. Bakel ◽  
George F. Vandegrift
Keyword(s):  
High Performance ◽  
Research Reactor ◽  
Reactor Fuel ◽  
Aqueous Processing ◽  
Performance Research

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

2018 ◽  
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.

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

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.

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

2014 ◽  
Vol 94 ◽  
pp. 43-54 ◽  
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).

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

2015 ◽  
Vol 1743 ◽  
Author(s):  
Jan-Fong Jue ◽  
Dennis D. Keiser ◽  
Tammy L. Trowbridge ◽  
Cynthia R. Breckenridge ◽  
Brady L. Mackowiak ◽  
...  
Keyword(s):  
Diffusion Couple ◽  
High Performance ◽  
Direct Interaction ◽  
Annealing Treatment ◽  
Rolling Process ◽  
Interaction Layer ◽  
Fuel Design ◽  
Enriched Uranium ◽  
Fuel Meat ◽  
Performance Research

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.

scholarly journals Design optimization methods for high-performance research reactor core design

2019 ◽  
Vol 352 ◽  
pp. 110167
Author(s):  
Benjamin R. Betzler ◽  
David Chandler ◽  
David H. Cook ◽  
Eva E. Davidson ◽  
Germina Ilas
Keyword(s):  
Design Optimization ◽  
High Performance ◽  
Research Reactor ◽  
Reactor Core ◽  
Optimization Methods ◽  
Performance Research
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