This presentation provides information on the experiments to measure the effect of Tantalum (Ta) on critical systems. This talk presents details on the Sandia Critical Experiments Program with the Seven Percent Critical Experiment (7uPCX) and the Burnup Credit Critical Experiment (BUCCX). The presentation highlights motivations, experiment design, and evaluations and publications.
This lecture is on the design of a Uranium Dioxide-Beryllium Oxide UO2-BeO Critical Experiment at Sandia. This presentation provides background info on the Annular Core Research Reactor (ACRR). Additionally, this presentation shows experimental and alternative designs and concludes with a sensitivity analysis.
Under IER-305, critical experiments will be done with and without molybdenum sleeves on 7uPCX fuel rods. New critical assembly hardware has been designed and procured to accomplish the experiments with the fuel supported by in a 1.55 cm triangular-pitched array.
A series of experiments will be performed to test the integral effects of molybdenum on the reactivity of a critical system. These experiments will use the 7uPCX assembly with the 1.55 cm triangular pitch grid plates. Molybdenum sleeves, consisting of 19.6 inch long 0.5-inch nominal outside diameter molybdenum tubes with 0.031-inch nominal wall thickness and centering hardware, will be placed on some of the fuel rods in the array. The purpose of this analysis is to examine two configurations of the 7uPCX using the 1.55 cm triangular pitch grid plates in fully-reflected approach-to-critical experiments with the number of fuel rods in the array as the approach parameter. This document presents the results of the analysis that was done to allow completion of the 7uPCX Configuration Checklist from Appendix A of SPRF-AP-005 [SNL 2020] for the cores noted above. The checklists for these cores are shown in Appendix A.
A new set of critical experiments exploring the temperature-dependence of the reactivity in a critical assembly is described. In the experiments, the temperature of the critical assembly will be varied to determine the temperature that produces the highest reactivity in the assembly. This temperature is the inversion point of the isothermal reactivity coefficient of the assembly. An analysis of relevant configurations is presented. Existing measurements are described and an analysis of these experiments presented. The overall experimental approach is described as are the modifications to the critical assembly needed to perform the experiments.
This presentation provides details regarding integral experiments at Sandia National Laboratory for fiscal year 2021. The experiments discussed are as follows: IER 230: Characterize the Thermal Capabilities of the 7uPCX; IER 304: Temperature Dependent Critical Benchmarks; IER 305: Critical Experiments with UO2 Rods and Molybdenum Foils; IER 306: Critical Experiments with UO2 Rods and Rhodium Foils ; IER 441: Epithermal HEX Lattices with SNL 7uPCX Fuel for Testing Nuclear Data; IER 452: Inversion Point of the Isothermal Reactivity Coefficient; and IER 523: Critical Experiments with ACRR UO2-BeO Fuel.
This presentation discusses activities related to the Nuclear Criticality Safety Program (NCSP) at Sandia National Laboratory in fiscal year 2021. This includes NCSP funding, integral experiment requests, integral experiment spending, highlights, and COVID-19 impacts.
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Transactions of the American Nuclear Society
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Estimation of the uncertainty in a critical experiment attributable to uncertainties in the measured experiment temperature is done by calculating the variation of the eigenvalue of a benchmark configuration as a function of temperature. In the low-enriched water-moderated critical experiments performed at Sandia, this is done by 1) estimating the effects of changing the water temperature while holding the UO2 fuel temperature constant, 2) estimating the effects of changing the UO2 temperature while holding the water temperature constant, and 3) combining the two results. This assumes that the two effects are separable. The results of such an analysis are nonintuitive and need experimental verification. Critical experiments are being planned at Sandia National Laboratories (Sandia) to measure the effect of temperature on critical systems and will serve to test the methods used in estimating the temperature effects in critical experiments.
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Transactions of the American Nuclear Society
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Under IER-209, critical experiments like those done under IER-208 will be performed. The difference between the two sets of experiments is that, while the IER-208 experiments were done in the 0.800 cm pitch 7uPCX hardware, the IER-209 experiments will be done in the 0.855 cm pitch 7uPCX hardware.
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Transactions of the American Nuclear Society
The intent of this work is to highlight the role of differential and integral data evaluation to address issues in connection with criticality safety applications. Demonstration is made by using, as an example, differential data measurements and evaluation, and the benchmark integral experiments for titanium. Energy-differential data are measured analyzed and evaluated to produce nuclear data libraries for criticality safety applications. Alternatively, integral experiments are performed at critical facilities, small experimental reactors, and play an important part in the validation of the differential nuclear data. The demonstration provided here for titanium gives its importance to criticality safety. Titanium is an effective neutron absorber that serves as baseline material for chemical separation in high-activity waste solutions in US. Titanium has not been considered for use in nuclear applications such as reactor design and analysis. Rather, it appears as a structural material that may be present in fuel cycle facilities or canisters for transport and disposition of nuclear waste. Criticality safety evaluations of systems in which titanium is present require an understanding of the nuclear data and its uncertainty.
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The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O2 fuel rods.
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ICNC 2015 - International Conference on Nuclear Criticality Safety
The Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories was designed to provide benchmark criticality and reactor physics data for water-moderated pin-fueled nuclear reactor cores in the 5 to 10 percent enrichment range. Approach-to-critical experiments were performed on nineteen partially-reflected arrays of pure water-moderated and -reflected fuel rods with a fuel-to-water volume ratio of 0.67. Those configurations are described and the results of the measurements are reported in this paper.
Transactions of the American Nuclear Society
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This presentation discusses the recent Sandia critical experiments, specifically the Seven Percent Critical Experiment (7uPCX) which is a Nuclear Energy Research Initiative (NERI) project. It also discusses why 7uPCX was used, how 7uPCX is operated and presents some 7uPCX results. The presentation concludes by discussing future plans for the critical experiments.
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This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.