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...
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.In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. This report also presents results of steady state neutronic analysis of an all-fresh LEU fueled core. Where possible, HEU and LEU calculations were performed for conditions equivalent to HEU experiments, which serves as a starting point for safety analyses for conversion of MITR-II from the use of HEU fuel to the use of UMo LEU fuel.
In awareness of the risk caused by the proliferation of nuclear materials, the international community initiated in 1978 the Reduced Enrichment for Research and Test Reactors (RERTR) Program [Ref. 1]. The goal of the RERTR program is to seek solutions to convert the reactors using High Enriched Uranium (HEU) fuel ( 235 U/U ≥ 20 wt. %) to the use of Low Enriched Uranium (LEU) Fuel ( 235 U/U < 20 wt. %). Among the 200 reactors worldwide currently in the scope of the program, about 70 have been converted or shut down prior to conversion. In the U.S., six high reactors remain to be converted [Ref. 2].One of the U.S. reactors is the High Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory (ORNL). This multipurpose reactor is mainly used for neutron scattering experiments and isotope production. It has two fuel elements made of involute-shaped fuel plates [Ref. 3].The ORNL Research Reactor Division (RRD) has prepared during FY2011 two series of calculations, one neutronic and one thermal-hydraulic (TH), in support of the conversion activities for HFIR. The neutronic calculations cover mainly the evaluation of the matrix of performance (cycle length and flux), the distribution of power in the fuel elements, kinetics parameters and reactivity coefficients [Ref. 4].As defined by the RRD procedures, the calculations have to be reviewed. Per internal decision, the RRD proposed an external reviewer, Argonne National Laboratory (ANL). The results of the Argonne neutronic review are presented in this report. The results of the TH review are presented in a separate report.The main conclusion of the ORNL reactor physics analyses is that the proposed reference HFIR LEU core (using the so-called monolithic UMo fuel [Ref. 5]) could maintain the current level of performance (obtained with a HEU core at 85MW) if it is operated at 100MW. However, to operate within the same safety limits as the current HEU core, the thermal hydraulic analyses show that the peaking of the power distribution occurring at the bottom of the fuel plates has to be reduced.The RRD has found a potential solution by reducing the fuel thickness at the bottom of the fuel plate. This would require addition of an axial grading to the already radially-graded fuel. Thus, the review has been focused on independently recalculating all of the ORNL calculations and making an as fair as possible comparison between the current 85MW HEU core and the reference 100MW LEU core design.ORNL and ANL have used the same computational tools to perform the work. The codes MCNP [Ref. 6] and ORIGEN [Ref. 7] have been used to perform the steady state and depletion calculations, respectively. The communication between the two codes has been carried out by the code VESTA [Ref. 8].Even though the code MCNP is able to handle complex geometries, it cannot model the involute shape. Nonetheless, ANL has developed an innovative methodology to model the involute shape with MCNP with an approximation that can be made as accurate as desired. Thus, ANL has been able to p...
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