Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

Abstract

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion 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), the MURR Facility, and the Nuclear Engineering Program -College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government.In the framework of 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 LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Al x dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo (U-10Mo).For the proposed LEU-fueled core, previous studies [1] have shown that in order to maintain the neutron flux at various crucial experimental locations within the reactor at the same level after the fuel conversion from the HEU to LEU, the steady-state operating power level must be increased from 10 MW to 12 MW. Hence, the accidental positive reactivity insertions, and the loss-of-coolant and loss-of-flow accidents for the LEU core were initiated from an initial steadystate power level of 12 MW. It is noted that all references to the core power level for MURR in this report are the thermal power output of the reactor (i.e., MWt).A search of the published literature and available technical reports was conducted to obtain appropriate values of thermal conductivity and volumetric heat capacity for the materials of 1) the irradiated and unirradiated HEU and LEU fuel meat, 2) the zirconium layers in the LEU fuel plates, 3) the 6061-aluminum of the fuel plate cladding, the reactor vessel walls, and primary system pipes, and 4) the oxide layer on the cladding surfaces of the irradiated fuel plates. This set of material properties was used in the analyses of the reactivity insertion accidents (RIA), loss-of-coolant accidents (LOCA), and loss-of-flow accidents (LOFA).The transient simulations of RIAs, LOCAs, and LOFAs are initiated from the conservative end of the normal operating band for reactor power, reactor primary coolant flow, and reactor core inlet temperature. The typical core state for the MURR was assumed in these simulations. This state has elements with a mixture of burnups, the flux trap is loaded with irradiation samples and experiments, and the control blades are banked in the...

University of Missouri Research Reactor LEU Fuel Element Flow Test Conceptual Design—Hydraulic Reactor Design Parameters

“…The MURR LEU preliminary Safety Analysis Report (SAR) [1], accident analyses report of MURR for LEU conversion [8], MURR irradiation demonstration element design parameter report [9], and MURR SAR [10] have been reviewed to collect the reactor design parameters relevant to the flow test. The results are summarized in Table 3.1.…”

“…(4.3) and (4.4) is calculated using the total flow area and wetted perimeter of an element. This simplified approach is also used in the previous safety analyses of MURR [5,8] for estimating the core flow rate of the analyses.…”

“…Assuming all fuel elements have the same channel gap thickness, the flow will be evenly distributed, and the flow rate per element can be calculated by dividing the total flow of 3750 gpm by the number of fuel elements (8), which is 468.75 gpm (rounded to 468.8 gpm throughout this report). However, the tolerance of the end channel gap thickness is relatively large, and thus, its effect on flow rate per element uncertainty should be evaluated, which is discussed in Section 4.2.1.…”

“…In MURR LEU safety analysis [5,8], a maximum of 0.008 inch channel gap reduction (0.004 inch for end channels) is used at the maximum burnup (180 MWd), and a linear relation between burnup and the channel gap reduction is assumed. The approach of 0.008 inch maximum channel gap reduction and linear relation with burnup is followed in this work.…”

“…The fuel element burnup amount used for the calculations presented here is assumed following the analysis for the MURR LEU equilibrium core [5,8]. Since the fuel elements of MURR are loaded in pairs, each pair of elements will have the same burnup level.…”