High Density Low Enriched Uranium Research Articles

Design and optimization of the new LEU MNSR for neutron radiography using thermal column to upgrade thermal flux

In this article, the preliminary feasibility of using the new miniature neutron source reactor (MNSR) with unique thermal column (TC), as a neutron source for neutron radiography (NR), is investigated. The new reactor core is designed based on the Isfahan MNSR with high density low-enriched uranium (LEU) UO2 fuel instead of high-enriched uranium (HEU) UAl4-Al fuel. The safety of the LEU reactor is studied by the calculation of important neutronics parameters such as core excess reactivity, central control rod worth, shut-down margin (SDM) and effective delayed neutron fraction. The vertical collimator dimensions, the materials to be employed in the collimator, and the collimator position are optimized using MCNP Monte Carlo code. The results showed that presence of the graphite thermal column increases thermal neutrons flux and reduces gamma-ray dose that can satisfy the requirements for neutron radiography. Therefore there is no need for extra gamma filters, fast neutron filters, and neutron moderators at the collimator. We emphasize here that the current designed LEU MNSR can be applied for NR and neutron activation analysis (NAA).

High-Fidelity Modeling and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies.The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuel meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. Calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008–2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation.This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Metallurgical considerations for the fabrication of low-enriched uranium dispersion targets with a high density for 99Mo production

It is necessary to develop high-uranium-density targets with low enriched uranium in order to improve 99Mo production efficiency. In order to investigate the use of high density low enriched uranium (LEU) dispersion targets using atomized U–Al alloy powder, U–10 wt% Al and U–20 wt% Al powders were fabricated by centrifugal atomization. Metallurgical reactions between uranium and aluminum during the hot working were investigated in order to minimize changes in the established chemical processes for 99Mo extraction. It is concluded that LEU based dispersion target plates with increased uranium density can be fabricated by controlling the chemical reactions between U and Al in atomized U–Al alloy powders.

Swelling of U(Mo) dispersion fuel under irradiation – Non-destructive analyses of the SELENIUM plates

Extensive fuel-matrix interactions leading to plate pillowing have caused a severe impediment on the development of a suitable high density low-enriched uranium dispersion fuel for high power applications in research reactors. Surface engineering of the U(Mo) kernel surfaces, where the interaction occurs, is put forward by SCK⋅CEN as a possible solution in the Surface Engineering of Low ENrIched Uranium Molybdenum fuel (SELENIUM) program. The project involved the construction of a sputter coater, the coating of U(Mo) kernels, the production of fuel plates, the irradiation and post-irradiation examination of 2 plates. The irradiation of 2 distinct (600nm Si and 1000nm ZrN coated) full size, flat fuel plates was performed in the BR2 reactor in 2012. The irradiation conditions were: 470W/cm2 peak Beginning Of Life (BOL) power, with a ∼70% 235U peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation.This paper reports the results and interpretation of the non-destructive post-irradiation examinations that were performed on these fuel plates and derives a law for the fuel swelling evolution with burnup for this fuel type. It further reports additional PIE results obtained on fuel plates irradiated in campaigns in the past in order to allow a complete comparison with all results obtained under similar conditions. The fuel swelling is shown to evolve linearly with the fission density, with an increase in swelling rate around 2.5×1021f/cm3, which is associated with the restructuring of the fuel. A further increase in swelling rate is observed at the highest burnups, which is discussed in this article.

Effects of high density dispersion fuel loading on the control rod worth of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the worth of the control rods of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U–Mo (9w/o) LEU fuels, having uranium densities of 6.57g/cm3, 7.74g/cm3 and 8.90g/cm3. Simulations were carried out to calculate the worth of the control rods at beginning of life. Three control rod materials were investigated: Ag–In–Cd alloy, B4C and Hf. Nuclear reactor analysis codes WIMS-D4 and CITATION were employed to carry out these calculations. It was seen that the control rod worth decreased as the uranium density of fuel increased from 6.57g/cm3 to 8.90g/cm3. The decrease was 10.58%, 9.39% and 9.74% for Ag–Cd–In alloy, B4C and Hf respectively. Even the increase in control rod thickness could not compensate for the decreased worth.

Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U–Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm 3, 7.74 gU/cm 3 and 8.57 gU/cm 3. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm 3 to 8.90 gU/cm 3. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

Effects of high density dispersion fuel loading on the dynamics of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the dynamics of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different properties affecting the reactor in different ways, fuels U–Mo (9w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to determine the reactor performance under reactivity insertion and loss of flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It is observed that during the fast reactivity insertion transient, the maximum reactor power is achieved and the energy released till the power reaches its maximum increases by 45% and 18.5%, respectively, as uranium density increases from 6.57 gU/cm 3 to 8.90 gU/cm 3. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved during the transient, by 27.7 K, 19.7 K and 7.9 K, respectively. The time required to reach the peak power decreases. During the slow reactivity insertion transient, the maximum reactor power achieved increases slightly by 0.3% as uranium density increases from 6.57 gU/cm 3 to 8.90 gU/cm 3 but the energy generated till the power reaches its maximum decreases by 5.7%. The temperatures of fuel, clad and coolant outlet remain almost the same for all types of fuels. During the loss of flow transients, no appreciable difference in the power and temperature profiles was observed and the graph plots overlapped each other.

Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U–Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.

Burnup study for Pakistan Research Reactor-1 utilizing high density low enriched uranium fuel

Burnup study for Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type MTR utilizing high density low enriched uranium fuel, was performed by using Fuel Cycle Analysis Program (FCAP). Existing equilibrium core of PARR-1, which is relatively economical but provides less neutron fluxes per unit power than the first equilibrium core, was formed by adding five more fuel elements in the first equilibrium core. This study shows that if the fuel loading is increased in the first equilibrium core of PARR-1 by replacing the fuel of density 3.28 gU/cm 3 by the fuel of density 4.00 gU/cm 3 then the new equilibrium core can provide 10% higher neutron fluxes at the irradiation sites and will also require 1.5 kg less fuel than that required for existing equilibrium core for one-year full power operation at 10 MW. The new core provides neutron fluxes at 13% lower cost and if the size of this core is further reduced by three fuel elements then this core can provide 20% higher thermal neutron flux at the central flux trap at 9% lower cost. A possible use of U-Mo (5 w/o Mo) fuel of density 8.5 gU/cm 3 in PARR-1 with an increase in existing water channel width from 2.1 to 2.45 mm (Ann. Nucl. Energy 32(1), 29–62) would provide up to 41% more thermal neutron flux at the central flux trap at 13% lower cost than the existing equilibrium core. The power peaking factors in these cores are similar to the power peaking factors of the existing equilibrium core and these cores are likely to operate within the safety constraints as defined for the existing equilibrium core of PARR-1.