Academic literature on the topic 'Light Enriched Uranium (LEU) Assembly'

One of the US Department of Energy’s Global Threat Reduction Initiative (GTRI) goals is the elimination of the use of high enriched uranium (HEU) for the production of the radioisotope molybdenum-99 (Mo-99). One strategy to achieve this goal is to use an irradiation target that utilizes a low-enriched uranium (LEU) foil. This paper considers an annular target geometry, where an LEU foil of open cross section is sandwiched between two concentric aluminum tubes. A recess is cut on the inner tube to hold the LEU foil and facilitate assembly. The target must contain the fission products until it can be opened and the LEU foil removed for further processing. The thermal contact resistance between the LEU foil and the aluminum tube cladding needs to be low enough to ensure that the LEU temperature does not exceed the operating temperature specified by the reactor safety case. Numerical models using the commercial finite element code Abaqus FEA are used to explore the potential for gaps opening between the LEU-foil and the aluminum tubes, and the stresses developed in the aluminum tubes during heat generation in the target during irradiation. Parametric studies are conducted using constructed analytic models to explore the impact of the ratio of heat transfer coefficients between the inner and the outer tubes and the inner-to-outer tube thickness ratio. It is concluded that the current annular target design is safe at high LEU heat generation rates and the use of these targets will likely not compromise the reactor safety.

As part of the Global Threat Reduction Initiative (GTRI) Reactor Conversion program, the fuel assembly at the University of Missouri Research Reactor (MURR) is undergoing a significant redesign. The proposed fuel structure is based on low-enriched uranium foils. The proposed aluminum-clad LEU foil fuel plates for the MURR core are significantly thinner than the currently used fuel plates. Further, the monolithic structure of the proposed fuel is fundamentally different than the current design based on powder metallurgy. Consequently, coolant flow reduction due to flow induced deformation of the proposed fuel plates is of concern. The goal of the current analysis is to estimate the amount of flow induced deformation of the proposed LEU-based fuel plates when subjected to coolant flow imbalance due to fuel plate assembly tolerances. Previous methods for assessing fuel plate deflection have relied heavily on analytic and experimental techniques. With the continued advancement of computational codes, new options are now available to assess structural stability. The current approach is to explicitly couple a commercial CFD code with a commercial FEM code. This paper will describe the convergence and stability criteria that were developed to obtain an accurate deflection solution. Time step management and pressure ramping strategies were effectively used as relaxation parameters to improve the computational stability. Additionally, mesh quality criterion were developed and are enforced during a simulation. Benchmarking of the numeric results to analytic calculations is also presented.

Safety margins are particularly tight in natural uranium-fuelled CANDU reactors which are refueled on-power. During on-power refueling, the insertion of xenon-free fresh fuel bundles into the reactor core affects the reactor’s excess reactivity in such a way that this could lead to temporary power derating. It is desirable from a fuel management perspective, and to maintain safety margins to eliminate this xenon-free effect and any other power ripples such as the subsequent plutonium reactivity peak. A redesign of the CANDU NU fuel bundle with an appropriate combination of elements, with some including neutron-absorbers, could well address the issue of the xenon-free initial portion of the bundle’s irradiation and also lower the plutonium-peak that occurs shortly thereafter. This may improve the fuel utilization (by further optimizing the fuelling strategy) and provide improved safety margins (by lowering the maximum channel and bundle powers). The use of neutron-absorbers in fuel design and manufacturing has been a regular practice in Light Water Reactor fuels for more than three decades. In CANDU applications, neutron absorbers have also been considered for the conceptual Advanced CANDU Reactor and the Low Void Reactivity fuel designs, for which the fissile content is made of low enriched uranium (LEU) or MOX fuels. The application to CANDU natural uranium (NU) fuel, however, especially as burnable poisons, is a relative novel approach. The reason for this is that the neutron economy in natural uranium-fuelled CANDU reactors is a prime concern, thus the addition of extra neutron absorbers is generally shunned. In our proposed application of burnable poisons to existing CANDU NU fuel design, because of low excess reactivity for NU fuel, the amount of neutron-absorber is expected to be restricted to small quantities and in a manner whereby the poison effect is restricted to the initial period of excess reactivity of a newly inserted fuel bundle. This implies that the impact on neutron economy would be relatively minimal, but the fuel performance would be significantly improved. Small amounts and appropriate mixtures of neutron absorbers were selected (approximately 500 mg of absorbers in a CANDU fuel bundle having a nominal weight of 24 kg). Preliminary results indicate that the fuelling transient and the subsequent reactivity peak can be lowered to improve the reactor’s operating margins. A parametric study using the Los Alamos National Laboratories’ MCNP 5 and Atomic Energy of Canada Limited’s WIMS-AECL 3.1 codes is presented in this paper. Details of this project and future work are also to be discussed.

Boron Neutron Capture Therapy (BNCT) is a kind of the targeted therapy with two element. It can kill the cancer cells while the effect on normal cells is very small, and it is suitable for the treatment of the various stage cancer so it will be the ideal radiotherapy for cancer treatment in the future. And Commercial Miniature Neutron Source Reactor (C-MNSR) was designed and constructed by CIAE, which is used for Neutron Activation Analysis (NAA), Training and teaching. The reactor with thermal power 27kW is an under-moderated reactor with pool-tank type, U-AL alloy with High Enriched Uranium (HEU) as fuel, light water as coolant and moderator, and metal beryllium as reflector. The fission heat produced by the reactor is removed by the natural circulation. Design C-MNSR with a epi-thermal neutron beam for BNCT is studied while the conversion from HEU to LEU (Low Enrichment Uranium) (235U percent≤20%) is carried on. As it has the advantages of MNSR safety, economy, easy operation and its application, and it can improve the epi-thermal neutron flux density and meet the requirements of BNCT. The fuel cage of C-MNSR with size of φ230×248mm in the reactor core, there are ton rows of 355lattices are concentrically arranged, the central lattice is reserved for central control rod, and four tie rods are uniformly arranged at the eighth row which link the upper and lower grid plates, the rest 350 fuel lattices are for fuel pins or dummies. The diameter of the fuel meat is 4.3mm, the height is 230mm, with Uranium enrichment is 17%; the diameter of the fuel element is 5.5mm, the height is 248mm. The frame design of the epithermal neutron beam is: Fluental material used as neutron moderation layer with its thickness is 50cm and its density is 2.85g/cm3; Cd with thickness of 0.1cm used as thermal neutron absorption layer, Lead with thickness of 10cm used as gamma ray shielding layer. And the neutron collimator parts is a composition of graphite, Cd and polythene with boron. The total length of the beam is 114.5cm, and the distance from the exit of the beam to the core is 130cm. The results show that the epithermal neutron flux density at the exit is 1.58 × 109n·cm-2·s-1 at full power of 27kW. and the fast neutron density at the exit is 5.45 × 107n · cm-2 · s-1 at full power. Fast neutron dose contamination (Df/ φepi) is 2.88 × 10−11Gy · cm2 · n−1 and gamma dose contamination (Dγ/φepi) 2.18× 10−14 Gy·cm2·n−1.

The ability of coated particles of enriched uranium dioxide (UO2) fuel to withstand high temperature and contain the fission products in the case of a loss of cooling event is a vital passive safety measures over traditional nuclear fuel active safety system to provide cooling. Hence, it is proposed in this study that Light Water Reactors (LWR) could be made safer by re-designing the fuel in the fuel assembly. A slender geometrical model with tube-to-particle diameter ratio N = 2.503 and porosity ε = 0.546 mimicking the proposed nuclear fuel in the cladding was numerically simulated. This is to investigate the heat transfer characteristics and flow distribution under buoyancy driven force expected in the cladding tube of the proposed nuclear fuel using a commercial code. Random packing of the particles is achieved by Discrete Element Method (DEM) simulation with the aid of Star CCM+. The temperature contour and velocity vector profile obtained can be said to be good illustration of anticipated heat transfer phenomenon to occur in the proposed fuel design. Similarly, heat transfer parameters such as particle-to-fluid heat transfer coefficient, Nusselt number, Grashof number and Rayleigh number were determined from simulated results and are presented. These parameters are of prime importance when analysing the heat transfer performance of a fixed bed reactor.