Engineering topics
Rivas, Andy
Publications and source records attributed to Rivas, Andy.
A system diagnostic and prognostic framework based on deep learning for advanced reactors
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Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design
We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.
Propagating neutronic uncertainties for FFTF LOFWOS Test #13
The safety evaluation conducted for licensing a Sodium-cooled Fast Reactor (SFR) may require modeling transients with best-estimate calculation tools that must first be validated against real-world measurements. To provide the community with a valuable benchmarking opportunity for validating SFR analysis tools and methods, the International Atomic Energy Agency (IAEA) initiated a coordinated research project (CRP) in 2018 for the analysis of the Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) Test #13. The impact of nuclear data uncertainties on neutronics parameters was previously investigated based on the COMMARA-2.0 covariance matrix. Since the transient simulation results are very sensitive to certain reactivity coefficients, it was decided to employ rigorous uncertainty propagation methods to quantify the impact of nuclear data uncertainties on the best-estimate predication of FFTF LOFWOS Test #13. The DAKOTA code is used to propagate neutronic uncertainties through SAS4A/SASSYS-1 transient simulations, while taking into account spatial and reaction-wise correlations within these uncertainties. This study shows that the remaining discrepancies observed between the Argonne National Laboratory (ANL) best-estimate results and the experimental measurements can be partly explained by the uncertainty associated with Gas Expansion Module (GEM) worth, which contributes the majority of the overall nuclear data uncertainty on the output from the FFTF LOFWOS Test #13 transient simulation. This study also confirmed the importance of including spatial and reaction-wise correlations of nuclear data uncertainties on feedback coefficients in the uncertainty propagation to avoid under-estimating their impact during the transient simulations.
Preliminary Benchmark Calculations of Spent Nuclear Fuel Isotopic Compositions Using BWR Assay Data
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Investigation of Polyurethane and Fire-Retardant Foams as a Radiological Contamination Fixatives - 20202
Fixation of radiological contamination in decommissioning and deactivation can reduce worker risk, and mitigate potentially hazardous conditions; however, nearly every marketed contamination fixative has been found to be flammable, a significant concern in radiological facilities. Coupled with this, industry fixatives are normally used as a thin coating which can present problems when attempting to stabilize irregular void spaces or areas that are difficult to access and ensure full coverage. To this end, Savannah River National Laboratory (SRNL), in conjunction with Florida International University (FIU), have begun investigating the applicability of polyurethane based foaming materials, which will expand to fill a given asymmetric volume and provide fire retardancy to ensure no release of contamination in the event of a facility fire. Commercial foams undergo exothermic reactions resulting in gaseous release that subsequently causes the curing foam material to expand. Heat generation during curing can potentially compromise rubber parts of an enclosure, a significant concern in nuclear facilities. To mitigate large body heat generation, SRNL has also examined multiple layering of the material to ensure compatibility of the curing material with a previously cured body of the same material. Environmental studies were also completed in order to determine the influence on polyurethane based foams curing process. These studies showed that nearly half of the materials tested delaminated from a cured body of the same material, a property not discussed in commercially available documentation on the products. Of note was the compatibility of the foaming material with the substrate, depending on the type material that the foam was cured in/on, significant delamination could occur. Mechanical testing via tensile pull tester, dynamic mechanical analysis (DMA), and thermogravimetric analysis (TGA) was also performed. (authors)
Concoction and Appraisal of Radiation Hardened Polyurethane Foams for DOE Decontamination and Decommissioning (D and D) Operations
Since World War 2, a number of nuclear facilities have been relocating their nuclear waste to long term storage, or final disposition. These facilities face monumental challenges in cleaning up the fixated transferrable radioactive waste that is left in these facilities after initial cleaning operations have ended. SRNL has identified a solution that is utilizing two-part epoxy foams with heavy metal additives for fixating and shielding transferable radioactive material that has accumulated in non uniform volumes. The purpose of this experiment is to make a library of data of physical characteristics of each foam and additive combination that will have data concerning how the material cures and performs in different environments with varying temperatures and how it will act and perform after being irradiated. Materials Being Tested: - Foam iT 7FR, Foam iT 23FR, 3M Fire Barrier foam, and Hilti CP620 foam; - Additives Bi, Bi2O{sub 3}, Na{sub 2}WO{sub 3}, and WO{sub 3} at 12.5% and 25%. What's being investigated: - Expansion data from each foam; - Radiation shielding data for each foam combination; - Peak temperatures for each foam. Utilized a FLIR thermal camera to find peak temperature and time it took the foam to reach it. No noticeable difference in volume or weight after blank foam samples sat in the environmental chamber for 24 hours except the 3M foam at the extreme relative humidity of 90% highlighted above. The more additives used in the foam, the more they were able to shield. The foams begin to saturate around 100% wt% where more additive used will not increase the shielding capability of the foam. Similar thermal dispersion in blanks and in samples with Bi additive. Time taken to peak temperature is very similar between the two. Similar peak temperatures in blanks and in Bi samples. Hilti had slightly elevated temperatures. All foams maintained their physical properties post environmental chamber except the 3M foam. With the increased sponginess from the high relative humidity's, 3M may be disqualified for further testing if its mechanical properties also suffer. While curing all foams had their peak temperature occur between 40 deg. C and 70 deg. C except the Hilti which occurred above 90 deg. C with and without additive. With this information, no foams would be disqualified. After radiation shielding studies, it was found that IT 14 with Bismuth and Bismuth Oxide Additives provided the best shielding of Co-60 with over 9% attenuation.
Characterization of Radiation Hardened Polyurethane Foams for DOE Decontamination and Decommissioning (D&D) Operations
The use of polyurethane foams has been under investigation to see if they can be used in DOE decontamination and decommissioning (D&D) operations, specifically to fixate and shield any remaining contamination that is left in any voided structure such as a glovebox. To fix this problem, SRNL attempted to characterize different commercial epoxy foams to evaluate their foam characteristics in different environments, temperature profiles, viscoelastic behavior, shielding capability, and any changes to these characteristics when different weight percent additives were included in the foam matrix. The use of these foams has not been realized yet because of excessive heat generation and flammability concerns during thecuring process.