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All figures (58)
Fig. 11. Relative difference (%) in burnup for a ROK assembly compared to operator-provided pin-by-pin burnups, using pin-by-pin burnup estimates based on (a) PDET gamma measurements in guide tubes and (b) operator data at the corner pins.
Fig. 10. A picture of PDET [31].
Fig. 1. Comparison of calculated (C) and experimental (E) 239Pu mass in different measured spent fuel samples. The calculations were performed using SCALE 6.1 and ENDF/B-VII libraries [9].
Fig. 2. Comparison of burnup codes in predicting the uranium and plutonium isotopic mass for the OECD/NEA benchmark phase 1-B, Case A [9]. (Note "C" stands for calculation and "E" for experiment.)
Fig. 4. Mass ratio of (a) 134Cs/137Cs and (b) 154Eu/137Cs with burnup.
Fig. 5. Gamma ratios vs. Burnup: (a) activity ratio of 134Cs/137Cs vs. burnup (b) activity ratio of (134Cs/137Cs)2/ (106Ru/137Cs) vs. burnup in three different fuel rods [27] (the symbols represent measured data points).
Fig. 16. Configuration of assembly NJ070G from TMI-1 [38]. Fuel rods O1, O11, O12, and O13 were extracted from the assembly for examinations after Cycle 10.
Fig. 17. Configuration of assembly NJ05YU from TMI-1 [38]. Fuel rods D5 and H6 were extracted from the assembly for examinations after Cycle 10.
Fig. 12. The average pin-by-pin burnup distribution (GWd/tU) in a ROK assembly provided by the reactor operator.
Fig. 13. The assembly pin-by-pin burnup distribution represented as the relative
Table 10. Burnup (GWd/tU) of the fuel samples from interior rods (D5 and H6) of assembly NJ05YU determined by different methods
Table 9. Burnup (GWd/tU) of the fuel samples from exterior rods (O1, O12, and O13) of assembly NJ070G determined by different methods
Fig. 36. The relative difference (%) in CIPN photon count rate due to different nuclide compositions generated by with- and without-neighbor models. Also shown is the relative difference (%) in 137Cs in each fuel rod between the two models.
Fig. 3. Comparison of operator-estimated and NDA measured burnup on a fuel rod [16].
Fig. 23. (a) Distribution of fast neutron flux in assembly NJ05YU and the neighbors in the energy group of [2.35, 2.48] MeV; (b) neighbor configuration of NJ05YU (duplicate of Fig. 21 (a)).
Fig. 34. Cross-sectional views of the CIPN instrument at two axial levels: (a) Z = -3 cm; (b) Z = 3 cm.
Fig. 22. Assembly NJ070G with its neighbor assemblies in Cycle 10 as modeled in TRITON.
Table 18. Summary of uncertainties in burnup of spent fuel determined using different methods
Fig. 6. Burnup vs. activity ratio at two different cooling times: (a) 134Cs/137Cs (b) (134Cs/137Cs)2/ (106Ru/137Cs).
Fig. 7. Mass ratio vs. burnup at two different power levels: (a) 134Cs/137Cs; (b) (134Cs/137Cs)2/(106Ru/137Cs).
Fig. 28. Sampler flowchart [43].
Fig. 29. The simplified 15×15 PWR spent fuel assembly as modeled in TRITON.
Table 1. Summary of PWR experimental assay data used for validation [9]
Fig. 27. Relative difference (%) in calculated nuclide concentrations of assembly NJ05YU between with- and without- neighbor models: (a) 244Cm after the first cycle; (b) 239Pu after the second cycle.
Fig. 35. The relative difference (%) in CIPN passive neutron count rate due to different nuclide compositions generated by with- and without-neighbor models. Also shown is the relative difference (%) in 244Cm in each fuel rod between the two models.
Table 6. Comparison of burnup determined from DA and NDA for Vandellos samples [16]
Fig. 33. Relative standard deviation of important fission products.
Table 5. List of conventional NDA instruments used to characterize or measure spent nuclear fuel
Fig. 26. (a) Calculated burnup (GWd/tU) in each fuel rod of assembly NJ070G after Cycle 10
Table 12. Average burnup of rods O1 and O12 in assembly NJ070G determined by different methods
Fig. 21. Neighbor assemblies of assembly NJ05YU (centered) in two different cycles: (a) Cycle 9; (b) Cycle 10.
Fig. 31. Distribution of calculated 239Pu mass results for 120 samples.
Fig. 30. Uncertainty in calculated 239Pu content as a function of burnup.
Fig. 18. Axial burnup profiles of fuel rod O1 and O12 measured by gamma scanning of 137Cs; converted to burnup using two different methods: 1) 148Nd measurements, 2) reference rod. Also shown are the destructive measurements of 148Nd at two axial locations.
Table 19. Summary of uncertainties in nuclide concentrations in spent fuel samplesa using different methods
Fig. 37. Relative difference of the passive gamma count rate of the samples from that of the reference case.
Fig. 8. Relative difference (%) between NDA and DA measured burnup vs. burnup in fuel samples from three different reactors.
Table 13. Relative difference (%) in predicted nuclide concentrations between the with- and without-neighbor models in the exterior rod O1 and the interior rod D5 of assembly NJ05YU
Table 14. Comparison of activity ratios of 154Eu/137Cs and 134Cs/137Cs of TMI fuel samples, from assembly NJ05YU NDA vs. calculation
Table 16. Relative difference (%) in neutron count rate of CIPN due to different burnup distributions within the assembly
Table 17. Relative difference (%) in gamma count rate of CIPN due to different burnup distributions within the assembly
Fig. 15. The calculated radial (a) and axial (b) plutonium concentration (g/tU) for the ROK assembly, with burnup gradient shown in Fig. 12, showing the large radial gradient and axial gradient near the ends of the fuel rods.
Table 11. Average burnup of rod D5 and H6 in assembly NJ05YU
Table 15. Relative differences (%) in nuclide concentrations calculated by four different modelsa compared to measurement for the TMI-1 sample O12S5 from assembly NJ070G
Fig. 32. Relative standard deviation of major actinides.
Fig. 9. Relative difference (%) between NDA and DA measured burnup vs. axial locations in spent
Table 3. Data required constructing a burnup model
Fig. 19. Axial burnup profile estimated by operator for assembly NJ070G, rod O1, and rod O12. Also shown is the burnup profile of rod O1 determined by gamma scan.
Table 8. Average burnups of rods O1 and O12 determined by different methods
Table 2. Summary of validation results [9]
Fig. 20. Neighbor assemblies of assembly NJ070G (centered) in Cycle 10. Previously irradiated assemblies are colored in yellow, while fresh ones are colored in white.
Fig. 39. Relative difference (%) of the net neutron count rate of the samples from that of the reference case.
Fig. 38. Relative difference of the passive neutron count rate of the samples from that of the reference case.
Table 4. Impacts of operator data uncertainties on nuclide concentrationsa,b
Table 7. Assembly design data for TMI-1 fuel
Fig. 14. Assembly axial burnup distribution (relative) measured by 137Cs gamma scan.
Fig. 24. (a) Calculated burnup (GWd/tU) in each fuel rod of assembly NJ05YU after the first cycle (Cycle 9) based on the with-neighbor model. (b) Relative difference (%) in burnup of each fuel rod of assembly NJ05YU after the first cycle between with- and without-neighbor models.
Fig. 25. (a) Calculated burnup (GWd/tU) in each fuel rod of assembly NJ05YU after the second cycle (Cycle 10) based on the with-neighbor model. (b) Relative difference (%) of burnup in each fuel rod of assembly NJ05YU after the second cycle between the with- and without-neighbor models.
Report
•
DOI
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Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project
[...]
Jianwei Hu
,
Ian C Gauld
,
James Banfield
,
Steven E. Skutnik
01 Mar 2014
-