Publications

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A method is described to test the effect of increased moderation on the 7uPCX critical arrays using the existing assembly hardware. The proposed experiments will allow the exploration of the assembly fuel-to-water ratio out to, and possibly beyond, optimum moderation in the assembly. A significant result reported below is that the total uncertainty in the benchmark k_eff in some of these experiments is reduced by about a factor of two compared to the uncertainties obtained in the fully-reflected experiments done to date.

The US Department of Energy Nuclear Energy Research Initiative (NERI) funded the Burnup Credit Critical Experiment (BUCCX) at Sandia National Laboratories. The BUCCX was designed to investigate the effect of fission product materials on critical systems. The BUCCX assembly was a water-moderated and -reflected array of Zircaloy-clad triangular-pitched U(4.31%)0₂ fuel elements. Some of the fuel elements could be opened to allow placement of experiment materials between the fuel pellets in the element. The ten BUCCX critical experiments reported here test the effect of the fission product rhodium on the assembly. The calculated reactivity worth of the rhodium in the experiments ranged from 0% for cases with no rhodium to a maximum of 3.5% of k_eff.

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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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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.

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 UO2 fuel rods. The uranium is enriched to 6.90% by mass. Sets of 36 titanium and aluminum experiment rods with the same nominal outside diameter as the fuel rods were fabricated and used as replacements for fuel rods in the array. The twenty-four 7uPCX critical experiments reported here compare the effects of the titanium and aluminum replacement rods on nearly critical fuel rod arrays. The fuel used in these experiments was fabricated using unirradiated UO₂ fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The nominal outside diameter of the fuel pellets is 0.207 in (0.52578 cm). The nominal outside diameter of the fuel rod cladding is 0.250 in (0.635 cm). The distance between the fuel rods in the square-pitched array is 0.315 in (0.8001 cm). This geometry gives a fuel-to-water volume ratio of 0.67 in the array. The twenty-four critical experiments in this series were performed in 2015 and 2016 at the Sandia Critical Experiments Facility. The first of the experiments had no replacement rods in the array and was intended to provide a baseline against which the experiments containing replacement rods could be compared. Eight critical experiments had titanium replacement rods in various numbers and arrangements near the center of the fuel array. Eight critical experiments had aluminum replacement rods in the same numbers and arrangements as in the eight experiments containing titanium experiment rods. In the final four experiments, fuel rods were removed from a central region of the array so that the pitch of the fuel rods in this part of the array was effectively doubled. This softened the neutron spectrum in the central part of the fuel array. Thirty-six replacement rods in different combinations of titanium and/or aluminum were placed in the interstices created in the center of the array. All twenty-four critical experiments are judged to be acceptable as benchmark experiments.

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This report examines proposed Seven Percent Critical Experiment (7uPCX) experiments with fuel arrays larger than would be critical when fully reflected. In these experiments, the reactivity of the assembly will be controlled by varying the moderator/reflector level in the core tank. The analysis uses two configurations, each completely filling the 45x45 fuel rod array with fuel rods and water holes, as representative examples of the proposed experiments. The proposed configurations are compared to the experiments documented in LEU-COMP-THERM-078 [Reference 1] and to fully-reflected experiments with the same fully-loaded fuel arrays that are poisoned with boron in the moderator. The conclusion is drawn that the proposed experiments can be performed with acceptably low uncertainties given a calibrated moderator/reflector level measurement system.

Under IER-206, the hardware for the Burnup Credit Critical Experiment will be removed from storage and placed in operation. This is the hardware we used in 2002 to perform the experiments that eventually documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments as LEU-COMP-THERM-079.

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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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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.

The Seven Percent Critical Experiment (7uPCX) was designed to provide benchmark criticality and reactor physics data for water-moderated pin-fueled nuclear reactor cores. The enrichment of the fuel was chosen to explore the enrichment range above the current 5% ceiling for US commercial pressurized water reactors. The experiment was part of the US Department of Energy (DOE) Nuclear Energy Research Initiative (NERI) Project 01-124 titled “Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel”. The NERI project was a collaboration between AREVA Federal Services, LLC; the University of Florida; Oak Ridge National Laboratory; and Sandia National Laboratories (SNL). The experiments at Sandia are currently supported by the DOE National Nuclear Security Administration Nuclear Criticality Safety Program. Two sets of benchmark experiments have been completed and documented as LEU-COMP-THERM-080 and LEU-COMP-THERM-078. Those experiments were done with the number of fuel rods in the fully-reflected array as the approach parameter. The experiments reported here are similar to those in LEU-COMP-THERM-080 except that the arrays are partially-reflected – the arrays were larger than would be possible with full reflection and the approach-to-critical experiments were done with the depth of the water in the critical assembly as the approach parameter. These experiments are reported as LEU-COMP-THERM-096.