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The thermal performance of the proposed low-enriched uranium (LEU) core for the University of Missouri Research Reactor (MURR) during steady-state operation is predicted. This conceptual design core, core CD35, is to replace the existing MURR Highly-Enriched Uranium (HEU) core. An earlier proposed LEU core design, the Feasibility Study Design (FSD) core, which was subsequently superseded due to issues of fuel plate manufacture, is also analyzed. Acceptable margins to both flow instability and critical heat flux (CHF) are demonstrated for both LEU cores. The designs of the HEU core and the two LEU cores are similar in that each has eight geometricallyidentical wedge-shaped fuel elements that are arranged in a circle to form an annulus. Each element has 23 or 24 parallel curved fuel plates depending on the element design, which are separated by thin curved rectangular coolant channels. There is an additional coolant channel outside the first fuel plate of each element and another one outside the last fuel plate. Downward flow through these coolant channels removes the power deposited in the core. Geometric tolerances and the clearances needed for insertion and removal of the elements from the reactor vessel are considered in the analysis. For each of the three cores, the core neutron physics analysis, which is provided in a separate report, considered 24 cases that together bound the most-limiting thermal-hydraulic state for each core. All 24 cases for each core were individually modeled in the thermal-hydraulic analysis. In each case all eight elements were modeled simultaneously. An acceptable LEU core must have sufficient margins to both flow instability and critical heat flux (CHF) events. Flow instability can occur in a reactor core that has parallel coolant channels. Added hydraulic resistance due to boiling in one channel can divert flow to another channel and cause the boiling channel to have a flow reduction excursion, or instability. When CHF occurs due to a flow excursion, the event is classified as a flow instability event rather than a CHF event. The CHF event is defined as one that is not caused by a flow reduction excursion. The operational safety margin to each event is predicted for all 24 cases of each of the three cores. The steady-state thermal-hydraulic analysis performed for the existing Safety Analysis Report (SAR) for the HEU core is well understood and can be easily duplicated. In 2011 some of the authors of the current report reviewed, replicated, and revised that analysis. ANL/RERTR/TM-12-37, Revision 1 was further analyzed with two other flow instability criteria. One of them assumes that flow instability occurs at the onset of significant voids, as predicted by Saha and Zuber 10. The other assumes that flow instability occurs when the minimum CHF ratio, based on the Bernath CHF correlation 6 , is 2.0. For the most limiting CHF case of each core, as predicted by the extended form of the Groeneveld 2006 CHF Table, the limiting channel was further analyzed with the Bernath CHF correlation....

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PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA, and has been used by the Reactor Conversion Program to determine the expected transient behavior of a large number of reactors. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature-and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available.

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