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 [1];  [1];  [1];  [1];  [1]
  1. Los Alamos National Laboratory
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Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
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Conference: International Conference on Nuclear Criticality ; 2015-09-14 - 2015-09-14 ; Charlotte, North Carolina, United States
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United States
Nuclear Physics & Radiation Physics(73)

Citation Formats

Hutchinson, Jesson D., Smith-Nelson, Mark A., Cutler, Theresa Elizabeth, Richard, Benoit Laurent, and Grove, Travis Justin. ESTIMATION OF UNCERTAINTIES FOR SUBCRITICAL BENCHMARK MEASUREMENTS. United States: N. p., 2015. Web.
Hutchinson, Jesson D., Smith-Nelson, Mark A., Cutler, Theresa Elizabeth, Richard, Benoit Laurent, & Grove, Travis Justin. ESTIMATION OF UNCERTAINTIES FOR SUBCRITICAL BENCHMARK MEASUREMENTS. United States.
Hutchinson, Jesson D., Smith-Nelson, Mark A., Cutler, Theresa Elizabeth, Richard, Benoit Laurent, and Grove, Travis Justin. 2015. "ESTIMATION OF UNCERTAINTIES FOR SUBCRITICAL BENCHMARK MEASUREMENTS". United States. doi:.
author = {Hutchinson, Jesson D. and Smith-Nelson, Mark A. and Cutler, Theresa Elizabeth and Richard, Benoit Laurent and Grove, Travis Justin},
abstractNote = {},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2015,
month = 1

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  • Subcritical experiments using californium source-driven noise analysis (CSDNA) and Feynman variance-to-mean methods were performed with an alpha-phase plutonium sphere reflected by nickel shells, up to a maximum thickness of 7.62 cm. Both methods provide means of determining the subcritical multiplication of a system containing nuclear material. A benchmark analysis of the experiments was performed for inclusion in the 2010 edition of the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Benchmark models have been developed that represent these subcritical experiments. An analysis of the computed eigenvalues and the uncertainty in the experiment and methods was performed. The eigenvalues computed usingmore » the CSDNA method were very close to those calculated using MCNP5; however, computed eigenvalues are used in the analysis of the CSDNA method. Independent calculations using KENO-VI provided similar eigenvalues to those determined using the CSDNA method and MCNP5. A slight trend with increasing nickel-reflector thickness was seen when comparing MCNP5 and KENO-VI results. For the 1.27-cm-thick configuration the MCNP eigenvalue was approximately 300 pcm greater. The calculated KENO eigenvalue was about 300 pcm greater for the 7.62-cm-thick configuration. The calculated results were approximately the same for a 5-cm-thick shell. The eigenvalues determined using the Feynman method are up to approximately 2.5% lower than those determined using either the CSDNA method or the Monte Carlo codes. The uncertainty in the results from either method was not large enough to account for the bias between the two experimental methods. An ongoing investigation is being performed to assess what potential uncertainties and/or biases exist that have yet to be properly accounted for. The dominant uncertainty in the CSDNA analysis was the uncertainty in selecting a neutron cross-section library for performing the analysis of the data. The uncertainty in the Feynman method was equally shared between the uncertainties in fitting the data to the Feynman equations and the neutron multiplicity of 239Pu. Material and geometry uncertainties in the benchmark experiment were generally much smaller than uncertainties in the analysis methods.« less
  • This work explores the constraints imposed on the /sup 235/U(n,f) standard (proposed ENDF/B version V) by information deduced from clean integral measurements and demonstrates how uncertainties in fission cross section standards propagate in an uncertainty analysis and interpretation of those experiments. The question of what a significant improvement in the accuracy of the /sup 235/U(n,f) standard would accomplish is addressed in the limited context of analyses of GODIVA and JEZEBEL measurements. The CSEWG integral benchmark results and uncertainties were updated in accordance with more recent information. Sensitivity coefficients were developed and used to estimate calculated results which should be obtainedmore » using the subsequent release of /sup 238/U(n,f), /sup 235/U(n,f), and /sup 239/Pu(n,f) at version V status. Covariance files were evaluated and processed for all important cross sections except inelastic scattering. Uncertainties due to the /sup 235/U(n,f) standard were estimated to comprise more than half of the calculated uncertainty for criticality and (/sup 28/sigma/sub f///sup 25/sigma/sub f/) /sub c/spectral index in JEZEBEL as well as GODIVA, although the JEZEBEL assembly contained no /sup 235/U. It is not possible at this time to predict criticality or (/sup 28/sigma/sub c///sup 28/sigma/sub f)/sub c/ to anywhere near the accuracy obtained by direct measurements, and therefore the integral results are significant to the analysis capability. Inclusion of integral information from GODIVA and JEZEBEL in an adjustment procedure was effective in reconciling all parameters other than (/sup 28/sigma/sub f///sup 25/sigma/sub f/)/sub c/ measurement in JEZEBEL. The adjustment procedure made changes of less than one standard deviation in the cross sections for /sup 235/U(n,f), /sup 235/U(n,..gamma..), /sup 238/U(n,f), /sup 238/U(n,..gamma..), and /sup 239/Pu(n,f) including an increase of approximately 1.5 percent for the /sup 235/U(n,f) cross section above 1.3 MeV.« less
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  • Computational Fluid Dynamics (CFD) evaluation of homogeneous and heterogeneous fuel models was performed as part of the Phase I calculations of the International Atomic Energy Agency (IAEA) Coordinate Research Program (CRP) on High Temperature Reactor (HTR) Uncertainties in Modeling (UAM). This study was focused on the nominal localized stand-alone fuel thermal response, as defined in Ex. I-3 and I-4 of the HTR UAM. The aim of the stand-alone thermal unit-cell simulation is to isolate the effect of material and boundary input uncertainties on a very simplified problem, before propagation of these uncertainties are performed in subsequent coupled neutronics/thermal fluids phasesmore » on the benchmark. In many of the previous studies for high temperature gas cooled reactors, the volume-averaged homogeneous mixture model of a single fuel compact has been applied. In the homogeneous model, the Tristructural Isotropic (TRISO) fuel particles in the fuel compact were not modeled directly and an effective thermal conductivity was employed for the thermo-physical properties of the fuel compact. On the contrary, in the heterogeneous model, the uranium carbide (UCO), inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers of the TRISO fuel particles are explicitly modeled. The fuel compact is modeled as a heterogeneous mixture of TRISO fuel kernels embedded in H-451 matrix graphite. In this study, a steady-state and transient CFD simulations were performed with both homogeneous and heterogeneous models to compare the thermal characteristics. The nominal values of the input parameters are used for this CFD analysis. In a future study, the effects of input uncertainties in the material properties and boundary parameters will be investigated and reported.« less