Finite Element Modeling of the TREAT (as Built) Reactor and a - PowerPoint PPT Presentation

Finite Element Modeling of the TREAT (as Built) Reactor and a Possible 20% Enriched Fuel TREAT PATRICK MCDANIEL NIMA FATHI CASSIANO DE OLIVEIRA

TREAT Modeling The Treat reactor was built in 1959 to test fuel for fast reactors It operated through 1994 An attempt is being made to restart TREAT as part of the DOE Advanced Fuels Initiative Originally using 93% enriched UO2, the restart may eventually require the use of < 20% enriched fuel. This provides an interesting neutronics problem to address the differences. As an aside, it seems interesting to point out that the original reactor was completed for a total cost of $1.46 M somewhat below the cost of the analyses that have been completed to address the restart.

TREAT Reactor

Plan View

Description The core support plate contained a grid of 19x19 locations that could contain fuel elements, control elements, thermocouple elements, or reflector elements. Each element was approximately 4 inches on a side and the fueled section was 4 feet long. Some elements had a 2 foot high slot in the middle that allowed instruments to view the central test section during a test. The fuel and reflector elements could be rearranged in a number of ways to obtain a critical configuration with the desired geometry.

TREAT Minimum Critical Core

TREAT Model All of the analysis performed here is based on the minimum critical core. Given the symmetry in the vertical dimension a 2-dimensional code was chosen for the analysis. FEMP2D is a finite element, p1, code that appears adequate for the analysis. Cross sections were derived from the 238 group criticality library for ENDFVII.0 provided with SCALE 6.1. The 238 group library was collapsed to 30 groups for most of the analysis – 20 fast and 10 thermal groups.

TREAT Model (2) Two adjustments were made to the original design as a result of ‘as built’ measurements. ◦ The carbon in the core material was estimated to be only 59% graphite and 41% free carbon ◦ The boron concentration was greater than anticipated and measured to be approximately 7 ppm as opposed to 1 ppm. The small core with control rods withdrawn had an excess reactivity of ~60 ih, or a ∆ k excess of 0.00157 We estimated the transverse buckling to be based on the fuel height plus 2 fast diffusion lengths on both ends to give a dimension of 157.5 cm. This gave a ∆ k excess of 0.00042. The exact ∆ k excess in our model required an effective vertical dimension of 166. 5 cm

Estimating the C/U-235 Ratio for the Restart K vs C/U-235 Ratio(sphere) K vs C/U-235 - 2D Model 1.1 1.0006 1.0005 1.05 1.0004 1 1.0003 0.95 1.0002 0 2000 4000 6000 8000 0 2000 4000 6000 8000 Sphere 2D Model To estimate the required C/U-235 ratio for the restart core to have the same size as the original, we started with a spherical model with a graphite reflector with an exactly critical original core. The graph on the left gives the results for the equivalent sphere. It contains no control rod elements. The red dots at the bottom are the results of the 2D calculation and the expanded plot on the right indicates that the ratio for C/U-235 of about 4000 is optimum.

Centerline Flux in the Reactor Static Fluxes at Centerline Static Fluxes at Centerline 6.01E-01 20% 93% Neutron Flux Per 1 Fission Neutron 1.00E-01 Neutron Flux per 1 Fission 5.01E-01 1.00E-03 4.01E-01 3.01E-01 1.00E-05 Neutron 2.01E-01 1.00E-07 1.01E-01 20% 93% 1.00E-09 1.00E-03 1.00E-05 1.00E-03 1.00E-01 1.00E-11 Neutron Energy (eV) 1.00E-05 1.00E-01 1.00E+03 1.00E+07 Neutron Energy (eV) The figures above compare the fluxes on the centerline of the small core. There does not appear to be a big difference over the fast range. Neither spectrum shows a fission hump and both are primarily 1/E spectra. The major difference occurs in the thermal range where the group around 0.01 ev shows a flux per fission neutron for the 20% enriched core is about 1/5 that of the original 93% enriched core.

As Built Temperature Feedback Coefficient Temperature Feedback Coefficient 0.00E+00 -5.00E-05 -1.00E-04 dk/dT -1.50E-04 -2.00E-04 -2.50E-04 -3.00E-04 300 500 700 900 1100 1300 Temperature o C Current Calculation Calculated ANL6173 1/sqrt(T) fit Measured

Temperature Feedback Comparison Temperature Feedback Coefficient Comparison 0 93% 20% -0.00005 dk/dT -0.0001 -0.00015 -0.0002 300 400 500 600 700 800 900 1000 1100 1200 Temperature o C

TREAT Restart Feedbacks Temperature Feedbacks 0 -0.00002 -0.00004 dk/dT -0.00006 -0.00008 -0.0001 -0.00012 Total Graphite Doppler 300 400 500 600 700 800 900 1000 1100 1200 Temperature ( o C)

U-238 Doppler Enhancement Doppler Feedback Enhancement Factor 1.8 1.7 Enhancement Factor 1.6 1.5 1.4 1.3 1.2 1.1 1 300 400 500 600 700 800 900 1000 1100 1200 Temperature o C

Transient Models Calculated TREAT As Built Lifetime 246 µ sec µ sec Calculated TREAT Restart Lifetime 185 α =(k-1)/l, l=(k-1)/ α Reported Lifetime 700 to 1000 µ sec Major discrepancy - still investigating Modeled 4 transients described in ANL-6173 Feedback coefficient reduced from 1.8E-4 to 1.3E-4 based on peak core temperature rather than an isothermal core Matched reactivity insertion, reduced lifetime by 25%, reduced temperature feed back by 45%, and increased core mass by 56.2 kg.

Transient Model Results Peak Power (MW) ∆ k ρ ($) Tau Measured PK Mod Restart(PK) 0.0095 1.319 0.31 54 62.8 140.3 0.0115 1.597 0.19 140 134 319 0.0155 2.153 0.105 380 399 1007.3 0.019 2.639 0.075 860 781 2025.1 Temperature Rise (K) ∆ k ρ ($) Tau Measured PK Mod Restart(PK) 0.0095 1.319 0.31 119 91.7 159 0.0115 1.597 0.19 145 110 191 0.0155 2.153 0.105 176 145 254 0.019 2.639 0.075 237 176 310

Conclusions A TREAT RESTART core is possible that will fit in the original footprint with 20% enriched fuel and a C/U-235 ratio of ~4000. The thermal flux on centerline will be significantly lower if no adjustments are made. The temperature feedback coefficient will be reduced by approximately 45%. Nominal transients will reach larger peak powers and greater core temperature rises.