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 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.

From High to Low Enriched Uranium Fuel in Research Reactors

2010 ◽  
Vol 73 ◽  
pp. 78-90 ◽  
Author(s):  
Sven van den Berghe ◽  
Ann Leenaers ◽  
Edgar Koonen ◽  
Leo Sannen
Keyword(s):  
Research Reactor ◽  
Current Status ◽  
Research Reactors ◽  
Dispersion Fuel ◽  
The Past ◽  
Uranium Fuel ◽  
The World ◽  
Enriched Uranium ◽  
The Moment ◽  
Global Threat

Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235U), low-density UAlx research reactor fuel with high-density, low enriched (<20% 235U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U3Si2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U3Si2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development.

High Performance Fine Grained Concrete with Content of Pumice

2020 ◽  
Vol 309 ◽  
pp. 21-25
Author(s):  
Vlastimil Bilek ◽  
Michal Pešata ◽  
Lukáš Procházka
Keyword(s):  
Mechanical Properties ◽  
Volcanic Rock ◽  
High Performance ◽  
Volcanic Glass ◽  
Volume Density ◽  
Point Of View ◽  
Fine Grained ◽  
Low Volume ◽  
Small Elements ◽  
High Absorption

Pumice is a volcanic rock that consists of a highly vesicular rough textured volcanic glass. It is very porous and shows a high absorption - it can contain a lot of water. Thanks to its low volume density, it is sometimes used as a light natural aggregate, especially for some small elements such as chimney blocks. The paper is focused on the development of concretes and mortars for these purposes. The optimum content of pumice was specified from the point of view of mechanical properties. Special attention was paid to absorbed water - it can be a source of self-curing of concrete.

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.

The Development of UMo/Al Dispersion Fuel

2011 ◽  
Vol 335-336 ◽  
pp. 1285-1292 ◽  
Author(s):  
Xiao Liu ◽  
Da Zhi Qian ◽  
Tie Cheng Lu
Keyword(s):  
High Performance ◽  
Aluminum Matrix ◽  
Dispersion Fuel ◽  
Existing Problems ◽  
Irradiation Test ◽  
Fuel Fabrication ◽  
The Matrix ◽  
Further Development ◽  
Performance Research ◽  
Irradiation Behavior

UMo/Al dispersion fuel is one of the prospective materials as a high uranium density fuel for high performance research reactors due to its excellent stability during irradiation. In this paper, An overview is provided of current development activities of UMo/Al dispersion fuel at abroad and home, including: the development reasons, this fuel fabrication technology, and the irradiation test. A comprehensive summary is given on the irradiation test, the existing problems and the solution recently obtained by the different countries. Early irradiation experiments with uranium alloys showed promise of acceptable irradiation behavior if these alloys could be maintained in their cubic γ-U crystal structure. The further development of this fuel was delayed due to an unacceptable volume expansion caused by UMo/Al interaction layer (IL) formation and a subsequent gross pore formation at the interface between UMo particles and matrix Al when severe irradiation conditions are reached. In order to alleviate or eliminate the swelling of UMo/Al dispersion fuel,several potential remedies are available to correct the swelling problems. These range from relatively minor changes to the fuel and matrix chemistry, to replacement of the aluminum matrix with another material, or to eliminate the matrix altogether. All of these variations are currently being investigated in the world.

From High-Enriched to Low-Enriched Uranium Fuel in Research Reactors

2010 ◽  
pp. 82-89
Author(s):  
S. Van Den Berghe ◽  
A. Leenaers ◽  
E. Koonen ◽  
F. Moons ◽  
L. Sannen
Keyword(s):  
Research Reactors ◽  
Uranium Fuel ◽  
Enriched Uranium

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.

High-density, low-enriched uranium fuel for nuclear research reactors

JOM ◽  
2003 ◽  
Vol 55 (9) ◽  
pp. 55-58 ◽  
Author(s):  
D. D. Keiser ◽  
S. L. Hayes ◽  
M. K. Meyer ◽  
C. R. Clark
Keyword(s):  
Nuclear Research ◽  
High Density ◽  
Research Reactors ◽  
Uranium Fuel ◽  
Enriched Uranium
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