A. Cuadra

Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

Analysis of Reactivity Insertion Accidents for the NIST Research Reactor Before and After Fuel Conversion

Reactivity insertion accidents have been analyzed for the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains highly enriched uranium fuel, and for a proposed equilibrium core with low-enriched uranium fuel. The time-dependent analysis of the primary system is performed with a RELAP5 model that includes the reactor vessel, primary coolant pump, heat exchanger, fuel element geometry, and flow channels for both the 6 inner and 24 outer fuel elements. Postprocessing of the simulation results has been conducted to evaluate minimum critical heat flux (CHF) ratio and minimum onset of flow instability (OFI) ratio using the Sudo-Kaminaga correlations and Saha-Zuber criteria, respectively. Evaluations are carried out for the control rod withdrawal start-up accident and the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that no damage to the fuel will occur and there is adequate margin to CHF and OFI because of sufficient coolant flow through the fuel channels and the negative reactivity insertion due to scram.

TRACE/PARCS Analysis of Anticipated Transient Without Scram with Instability for a MELLLA+ BWR/5

Abstract A TRACE/PARCS model has been developed to analyze anticipated transient without scram (ATWS) events for a boiling water reactor (BWR) operating in the maximum extended load line limit analysis-plus (MELLLA+) expanded operating domain. The MELLLA+ domain expands the allowable operation in the power/flow map of a BWR to low flow rates at high-power conditions. Such operation exacerbates the likelihood of large-amplitude power/flow oscillations during certain ATWS scenarios. The analysis shows that large-amplitude power/flow oscillations, both core-wide and out-of-phase, arise following the establishment of natural-circulation flow in the reactor pressure vessel after the trip of the recirculation pumps and an increase in core inlet subcooling. The analysis also indicates a mechanism by which the fuel may experience heatup that could result in localized fuel damage. TRACE predicts that heatup will occur when the cladding surface temperature exceeds the minimum stable film boiling temperature after periodic cycles of dryout and rewet, and the fuel becomes locked into a boiling-film regime. Further, the analysis demonstrates the effectiveness of the simulated manual operator actions to suppress the instability.

ANALYSIS OF LOSS-OF-FLOW ACCIDENTS FOR THE NIST RESEARCH REACTOR WITH FUEL CONVERSION FROM HEU TO LEU

Abstract A program is underway to convert the current high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel in the 20-MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology. A RELAP5 model has been developed to analyze postulated accidents in the NBSR with the present HEU fuel and a proposed LEU fuel. The model includes the reactor vessel, primary pumps, shutdown pumps, various valves, heat exchangers, and average and hottest fuel elements and flow channels in the region where flow enters through an inner plenum (6 fuel elements) and a region where flow enters through an outer plenum (24 elements). The equilibrium cycle power distributions in the fuel elements were determined based on three-dimensional Monte Carlo neutron transport calculations performed with the MCNPX code. In this paper we discuss safety analyses conducted for the loss-of-flow accidents resulting from either loss of electrical power or inadvertent throttling of flow control valves at the inlets to the inner and outer plena. The analysis shows that the fuel conversion will not lead to significant changes in the safety analysis and that the calculated maximum clad temperatures, minimum critical heat flux ratios, and minimum onset of flow instability ratios assure that there is adequate margin to fuel failure.