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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. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem = 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines. [ABSTRACT FROM AUTHOR]

The Massachusetts Institute of Technology Reactor (MITR) is a 6 MW tank-type reactor with a compact core design. It is moderated and cooled by light-water and has a heavy-water reflector. The MITR is a unique test bed for a wide range of nuclear material and fuel research. In one of the recent fuel cycles, detectors found elevated neutron levels on the reactor top lid during the operation with the Water-loop facility installed. A preliminary investigation theorized that the neutrons were streaming out via a thin CO 2 gas gap, which is used as a thermal isolation space for the in-core experiment. The current study employs neutron transport simulations to investigate the physical interpretation of the observed phenomenon. The state-of-the-art general purpose Monte Carlo code MCNP5 is adopted as the computational tool. The calculation results identify the typical characteristics of the streaming effect by quantifying the fraction of upward-streaming neutrons. Efforts have also been made to explore feasible engineering options for mitigating this neutron streaming from the Water-loop facility. It is concluded that a borated guide tube is not sufficient to reduce the neutron level on the reactor top lid; however, a stepped thimble design is effective in reducing the neutron source contribution. [ABSTRACT FROM AUTHOR]

Abstract: The Massachusetts Institute of Technology Research Reactor (MITR), with thermal power of 6 MW, is one of the five high-performance research reactors operated in the U.S. An in-house program MCODE-FM (MCNP-ORIGEN Coupled Depletion Program – Fuel Management) as the reference code supporting routine operation is currently developed by the MITR staff. Adopting Monte Carlo methods enables exact modeling of the MITR core geometry with use of continuous-energy nuclear data. A criticality search algorithm to track control blade movement is implemented in MCODE-FM. The code also features automation of input file generation, data manipulation, and post-processing of output data for fuel cycle analysis. Some verification and validation runs have been carried out by comparisons with the multiplication factor and the reactivity worth induced by control blades. In this study, further validation runs are performed, with two sets of measured data: 1) reactivity effects of fission product poisoning and fuel depletion and 2) reaction rate based thermal and fast neutron flux. Fission product poisoning effects calculated by MCODE-FM are in good agreement with the measured data obtained during the reactor start-up process (within 10%). The trend of peak xenon reactivity after the reactor shutdown is predicted accurately as well. As compared to the neutron activation experiment, MCODE-FM is found to slightly over-estimate the thermal flux with less than 10% discrepancy and the fast neutron flux with less than 20% discrepancy. [Copyright &y& Elsevier]

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to power light water reactors (LWRs) in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion of U.S. high performance research reactors (USHPRRs) like the MITR. The Preliminary Safety Analysis Report (PSAR) for the MITR LEU conversion has been submitted and currently under review by the U.S. Nuclear Regulatory Commission (NRC). A transition core plan, from 22 fresh LEU fuel elements (i.e., beginning-of-life) gradually to 24 of them arranged in an equilibrium configuration, is expected to serve as an appendix chapter in the final SAR. The current study presents the fuel cycle development, which eventually leads to the transition core plan. The results confirm the equilibrium state, where both the shim bank movement (i.e., core reactivity) as well as the content of fissile materials stabilize, can be achieved by fixed-pattern fuel management. Fission density has been evaluated for a number of fully discharged LEU fuel elements, using a best-estimate approach. The fuel cycle calculations also generate power profiles at each core state. A steady-state thermal-hydraulic safety analysis has thus been performed, where onset of nucleate boiling (ONB) is considered as the safety criterion. The results confirm significant margin to ONB at all analyzed transition and equilibrium fuel cycle states. [ABSTRACT FROM AUTHOR]

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. In the framework of non-proliferation policy, research and test reactors have started a program to convert highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. A new type of LEU fuel based on a high-density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion. A transition core plan, from 22 fresh LEU fuel elements to 24 in an equilibrium state, is proposed. This study evaluates the thermal-hydraulic safety margins of each transition cycle and state (i.e. beginning, middle and end). The STAT7 and RELAP5 codes are used in the study. STAT7 provides an integrated platform for statistical propagation of uncertainties in determining steady-state thermal-hydraulic margins. A RELAP5 model was created to verify MITR's STAT7 model, and provide transient analyses. The steady-state analyses determined that, during the entire transition from a fresh to equilibrium core, the Limiting System Safety Settings (LSSS) has a 10.2% margin to the minimum power at which onset of nucleate boiling (ONB) is precluded at 3 σ confidence. The loss of flow (LOF) transient, which is the most limiting anticipated transient for the MITR, was simulated. With a conservative initial state of the reactor (operating at μ + 3 σ power level and LSSS mass flow rate), the maximum calculated centerline fuel temperature was 115 °C, which is a significantly lower value than the blistering limit of 350 °C. The cladding wall temperatures did not exceed the ONB temperature throughout the transient. Therefore, during LOF, no nucleate boiling is expected, precluding any critical heat flux trigger. This study indicates that the proposed LEU transition core specifications have significant margin to thermal-hydraulic limits, during steady-state operation and the LOF transient. [ABSTRACT FROM AUTHOR]

This paper presents an experimental demonstration of a data-driven control system (DCS) designed for the MIT Graphite Exponential Pile (MGEP). The DCS aims to regulate the neutron flux profile such that symmetry is preserved. Neutron flux perturbations are introduced into the MGEP to test the DCS's capabilities by the movement of an initiating control rod (ICR). To realize this functionality, a control system that relies on an artificial neural network (ANN) was developed, and then demonstrated on the MGEP. A Helium-3 (3He) neutron detector and dual control rods, including their moving mechanisms, were fabricated. The perturbed flux profile was monitored by the moving neutron detector. The prediction accuracy of the neural network (NN) was examined and the DCS response was presented. Our results show that neural network regression model trained by experimental data can achieve a prediction error of less than 2.5 cm with a 95% confidence interval. The demonstration experiment also shows that a perturbation of the ICR can be captured by the control system and flux symmetry can be maintained within 1% after the response of the responding control rod (RCR). [ABSTRACT FROM AUTHOR]