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An experimental and computational effort was undertaken in order to evaluate the capability of the fluid-structure interaction (FSI) simulation tools to describe the deflection of a Missouri University Research Reactor (MURR) fuel element plate redesigned for conversion to lowenriched uranium (LEU) fuel due to hydrodynamic forces. Experiments involving both flat plates and curved plates were conducted in a water flow test loop located at the University of Missouri (MU), at conditions and geometries that can be related to the MURR LEU fuel element. A wider channel gap on one side of the test plate, and a narrower on the other represent the differences that could be encountered in a MURR element due to allowed fabrication variability. The difference in the channel gaps leads to a pressure differential across the plate, leading to plate deflection. The induced plate deflection the pressure difference induces in the plate was measured at specified locations using a laser measurement technique. High fidelity 3-D simulations of the experiments were performed at MU using the computational fluid dynamics code STAR-CCM+ coupled with the structural mechanics code ABAQUS. Independent simulations of the experiments were performed at Argonne National Laboratory (ANL) using the STAR-CCM+ code and its built-in structural mechanics solver. The simulation results obtained at MU and ANL were compared with the corresponding measured plate deflections. ANL/RTR/TM-16/9Evaluation of Thin Plate Hydrodynamic Stability through a Combined Numerical Modeling and Experimental Effort ii velocity as expected and the maximum deflection for multiple experiments performed with an average flow velocity near 4.3 m/s was found to range between 5.5-7.0 mil. Due to the apparatus the plate deflection could only be measured along an azimuthally-centered line. Therefore simulations of the curved-plate experiments were performed with two models: a model that assumes that the channels are azimuthally uniform and a model that assumes azimuthally varying channels. The results obtained with the azimuthally uniform model agree well with the measured results. The azimuthally uniform model predicts a deflection of 5.6 mil for the average coolant velocity of 4.25 m/s, which under-estimated by 8% the 6.06 mil fit of the measured deflections at this fluid velocity. The results obtained with the azimuthally varying model provide a measure of the sensitivity of the deflection results to the geometry of the curved plate. The azimuthally varying model over-estimates the measured plate deflection at the leading edge by approximately 80% for an average coolant velocity of 4.0 m/s. The measured values, even at their upper 95% confidence limit of the best fit, remain bounded by the azimuthally non-uniform model.Based on the modeling of experiments there are certain points to consider relative to the expected stability of a prototypic LEU MURR fuel plate. The comparison of FSI simulation results with the measured experimental plate deflections shows that when the plate and...

The University of Missouri (MU) has been working in conjunction with Argonne National Laboratory (ANL) in the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program to support conversion of the University of Missouri-Columbia Research Reactor (MURR®) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. MURR is one of five U.S. High Performance Research Reactors (USHPRR) plus a critical facility that plans to convert to the use of LEU fuel. ANL/RTR/TM-18/16 Safety Analysis of the Mo-99 Production Upgrade to the University of Missouri Research Reactor (MURR) with Highly Enriched and Low-Enriched Uranium Fuel ii ANL/RTR/TM-18/16 Safety Analysis of the Mo-99 Production Upgrade to the University of Missouri Research Reactor (MURR) with Highly Enriched and Low-Enriched Uranium Fuel iv installed. Thus, the upgrade causes a change in the location of the maximum HEU fuel temperature resulting from the limiting LOFA, which is consistent with the shift in the core power distribution that occurs due to the upgrade. For the LEU core, the minimum margin to the fuel temperature safety limit for the limiting LOFA with the 2017-RBM-99 upgrade is 232 °C and occurs in an EOL fuel plate. This is the same margin as was predicted for the limiting LOFA for the PSAR analysis. The location of the minimum margin moves from plate 23 of the fuel element in core position 8 in the PSAR analysis to plate 23 of the fuel element in core position 4 with the upgrade. With more margin than the other scenarios, all LOFA transient results are considered to have very adequate margins to the burnupdependent fuel temperature safety limit. Because there is no change to the HEU or LEU nominal fuel element loadings and an overall negligible effect on the fuel burnup as a result of the 2017-RBM-99 upgrade, the radiological consequences of the maximum hypothetical accident and fuel handling accident were not evaluated in this work. As found in the PSAR analysis, for any event where a release is considered in the current HEU SAR, the radiological consequences after conversion remain lower than the regulatory limits. In summary, the Mo-99 production upgrade to MURR with the 2017-RBM-99 device has been found to cause a shift in the core power distribution for both HEU and LEU cores relative to the PSAR analysis. The predicted steady-state margins to flow instability with the upgrade are adequate for both HEU and LEU cores. The margins with the upgrade are lower than those predicted in the PSAR analysis for both HEU and LEU, but the margin for LEU is larger than that for HEU. For postulated transient accidents, the upgrade causes a slight decrease in the minimum margin to a fuel temperature safety limit based on the fuel blister threshold temperature. There is a 9 °C reduction in the safety margin for HEU for the most limiting accident evaluated under conditions specified in NUREG-1537. For a reference LEU core, the minimum safety margin for the most limiting transient accident with ...

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel.

60439. For information about Argonne and its pioneering science and technology programs, see www.anl.gov.

EXECUTIVE SUMMARYThis report contains the results of reactor fuel element design for conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly enriched uranium (HEU) fuel to the use of low enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the Massachusetts Institute of Technology Nuclear Reactor Laboratory. The core conversion to LEU is being performed with financial support of the U. S. government.The goal of this work was to design an MITR LEU fuel element that could safely replace the current MITR HEU fuel element and maintain mission performance while requiring minimal, if any, changes to the reactor system. As a means to accomplish this, neutronic and steady-state thermal hydraulic performance of the MITR was analyzed with various LEU fuel element designs. The evaluation included the impact of assumed manufacturing tolerances and other uncertainties in reactor parameters.Documents that were reviewed as bases for the design and safety evaluations were the MITR design drawings and historic analyses of the facility. All of the information and data needed to construct the reactor models and perform the analyses were provided by MITR. The current HEU fuel element has 15 plates that are 0.080 inch thick with 0.010 inch deep grooves along the length of the plate. These grooves serve as fins to increase heat transfer area to the coolant. The HEU aluminide fuel contains uranium with a 235 U enrichment of 93 wt%, and is 0.030 inch thick in each plate. The Al-6061 aluminum cladding at the base of the grooves on the HEU plates is 0.015 inch thick.Prior LEU element design analyses with high-density monolithic alloy fuel have obtained equivalent performance and fuel cycle with an 18-plate element with 0.020 inch thick fuel and 0.010 inch cladding thickness (at the base of the fins). Whereas this prior MITR LEU design was based upon cladding thickness of 0.010 inch, recent manufacturing development experience has led to a re-evaluation of the minimum cladding thickness for reliable fabrication. These core and element design activities were undertaken to determine if additional cladding thickness could be incorporated into an MITR LEU element design. Since increased cladding thickness would displace water and degrade core reactivity, removal of the fins was also explored as a goal of the work. Removal of the fins would not only increase water to metal ratio in the core, but would also improve the economics of an MITR LEU element by eliminating this fabrication step, which is unique to MITR fuel design among U.S. high performance research reactors which refuel with HEU. In order to compensate for the loss of heat transfer area, an increased core coolant flow rate has been considered, and distinct fuel thicknesses were introduced in the outer plates of each element to limit heat flux peaking.The proposed LEU fuel element designs have the same over...

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60439. For information about Argonne and its pioneering science and technology programs, see www.anl.gov.