Highly Enriched Uranium Metal Annuli and Cylinders with Polyethylene Reflectors and/or Internal Polyethylene Moderator
- Georgia Institute of Technology, Atlanta, GA (United States)
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
A variety of critical experiments were constructed of enriched uranium metal during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, experiments of uranium metal annuli with and without polyethylene reflectors and with the central void region either empty or filled with polyethylene were evaluated under ICSBEP Identifier HEU-MET-FAST-076. The outer diameter of the uranium annuli varied from 9 to 15 inches in two-inch increments. In addition, there were uranium metal cylinders with diameters varying from 7 to 15 inches with complete reflection and reflection on one flat surface to simulate floor reflection. Most of the experiments were performed between February 1964 and April 1964. Five partially reflected (reflected on the top only) experiments were assembled in November 1967, but are judged by the evaluators not to be of benchmark quality. Twenty-four of the twenty-five experiments have been determined to have fast spectra. The only exception has a mixed spectrum. Analyses were performed in which uncertainty associated with five different parameters associated with the uranium parts and three associated with the polyethylene parts was evaluated. Included were uranium mass, height, diameter, isotopic content, and impurity content and polyethylene mass, diameter, and impurity content. There were additional uncertainties associated with assembly alignment, support structure, and the value for ßeff. In addition to the idealizations made by the experimenters (removal of a diaphragm), a few simplifications were also made to the benchmark models that resulted in a small bias and additional uncertainty. Simplifications included omission of the support structure, possible surrounding equipment, and the walls, floor, and ceiling of the experimental cell. Bias values that result from these simplifications were determined and associated uncertainty in the bias values were included in the overall uncertainty in benchmark keff values. Bias values ranged from 0.0002 ?k to 0.0093 ?k below the experimental value. Overall uncertainties range from ? 0.0002 to ? 0.0011. Major contributors to the overall uncertainty include uncertainty in the support structure and the polyethylene parts. A comparison of experimental, benchmark-model, and MCNP-model keff values is shown in Figure 1. The experimental keff values are derived from the original reactivities reported by the principal experimentalist. The benchmark-model keff values are the experimental keff values adjusted to account for biases that were introduced by removing the support structure and surroundings. The MCNP-model keff values are simply the values found from MCNP calculations using the benchmark specifications and ENDF/B-VI cross-section data. Figure 1. Comparison of Experimental, Benchmark-Model and MCNP-Model keff value. Calculated results for most of the experiments are
- Research Organization:
- Georgia Institute of Technology, Atlanta, GA (United States); Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
- DOE Contract Number:
- AC07-99ID-13727
- OSTI ID:
- 912458
- Report Number(s):
- INL/CON-06-11658; TRN: US0800407
- Resource Relation:
- Conference: 8. International Conference on Nuclear Criticality Safety (ICNC 2007), Evaluation of Nuclear Data, St. Petersburg (Russia), 28 May - 1 Jun 2007; Related Information: https://www.ans.org/meetings/view-6/
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
ALIGNMENT
BENCHMARKS
CALCULATION METHODS
CASTING
CRITICALITY
CROSS SECTIONS
ENRICHED URANIUM
GRAPHITE
HIGHLY ENRICHED URANIUM
MODERATORS
OR-CEF REACTOR
POLYETHYLENES
REFLECTION
SAFETY
SPECIFICATIONS
SPECTRA
STORAGE
URANIUM
VERIFICATION
Y-12 PLANT
Nuclear Criticality Safety Program (NCSP)
Uranium Annuli
ICSBEP
Polyethylene Reflectors
Uranium Metal Cylinder
Monte Carlo N-Particle Transport (MCNP)
Evaluated Nuclear Data File (ENDF)