Evaluation of Chemical Assay Data

In burnup credit applications to spent fuel management systems the isotopic inventory of the irradiated fuel is required as necessary and direct input to the criticality analysis of these systems. Therefore, as regards depletion code validation with respect to burnup credit applications, the attention is mainly focused on comparisons between measured and calculated isotopic concentrations.

2.2.1.Actinides and Burnup Indicators

The SAV system is verified against numerous chemical assay data. Most of these data refer, however, to actinides and burnup indicators (Nd-148 mostly). Figures 2 through 10 show a few examples obtained from assays of UO2, reprocessed UO2, MOX, and reprocessed MOX fuel assemblies. There are lots of more verification results including evaluations of assays on burnable absorber bearing fuel rods as well as evaluations of assays on nuclide concentrations as a function of the pellet radius [1]. All these results show:

1. that the measured and calculated isotopic ratios or concentrations are mostly in agreement within the experimental error bounds,

2. that SAV does not underestimate the concentrations of the fissile isotopes, and

3. that SAV tends to slightly underestimate the fertile actinides U-236, Pu-242 and sometimes also Pu-240.

An underestimation of the fertile isotopes U-236, Pu-242 and Pu-240 in particular results in an overestimation of the neutron multiplication factor of a thermal system. Usually the systems which require burnup credit are thermal systems.

FIG. 2. Fuel Nuclide Concentrations in UO2 Rods: Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Isotopic Fractions.

FIG. 3. Fuel Nuclide Concentrations in UO2 Rods: Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Isotopic Fractions.

FIG. 4. Fuel Nuclide Concentrations in UO2 Rods: Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Isotopic Fractions.

FIG. 5. Fuel Nuclide Concentrations in UO2 Rods: Verification of the Siemens Standard Design Procedure SAV: : Mass Ratio Nd-148/U-238 as a Function of the Ratio U-235/U (Curve a) (for the Transformation of the Nd-148 Scale into the More Convenient Burnup Scale the Mass Ratio Nd-148/U-238 is Calculated as a Function of Burnup as Shown by the Curve b).

FIG. 6. Fuel Nuclide Concentrations in Reprocessed UO2: Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Isotopic Fractions.

b a

FIG. 7. Fuel Nuclide Concentrations in Reprocessed UO2: Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Isotopic Fractions.

FIG. 8. Fuel Nuclide Concentrations in Reprocessed UO2: Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Isotopic Fractions.

FIG. 9. Fuel Nuclide Concentrations in First Generation Recycle MOX fuel (NPP

Obrigheim): Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Concentrations (Th:= Calculated, M:= Measured)

FIG. 10. Fuel Nuclide Concentrations in Second Generation Recycle MOX fuel (NPP Obrigheim): Verification of the Siemens Standard Design Procedure SAV: Comparison of Measured and Calculated Concentrations (Rechnung:= Calculation, Messung:= Measurement).

-30

E:= Measured mass ratio of the isotope specified to U-238, C:= Calculated mass ratio of the isotope specified to U-238 : Differences between Measured (E) and Calculated (C) Isotopic Mass Ratios.

-20

-15

-10

-50

510

15

20

25

30 U-234U-235U-236U-238Pu-238Pu-239Pu-240Pu-241Pu-242 Actinide

(C-E) / C / %

Sample 1: 3.5 wt.-%, 59.6 MWd/kg U Sample 2: 3.8 wt.-%, 54.5 MWd/kg U E) and Calculated (C) Isotopic Mass

-200

Sample 1: 3.5 wt.-%, 59.6 MWd/kg U Sample 2: 3.8 wt.-%, 54.5 MWd/kg U 13. Evaluation of Chemical Assay on Fuel Samples from NPP GKN II : Differences between Measured (E) and Calculated (C) Isotopic Mass

-2000

2.0 wt.-%, 9.9 MWd/kg 2.5 wt.-%, 17.4 MWd/kg 3.0 wt.-%, 23.9 MWd/kg 3.5 wt.-%, 30.4 MWd/kg 4.0 wt.-%, 36.7 MWd/kg 4.5 wt.-%, 42.6 MWd/kg 2 Fuel in Storage Racks of the Region II Type: Reactivity Worth of Fission Products at the Time of Reactor Uncertainty: ± 50 pcm.) (Initial Enrichments and Burnups Specified Correspond of the Loading Curve of the Region II Racks).

FIG. 15. Storage of Spent KONVOI UO2 Fuel in Storage Racks of the Region II Type:

Relative Changes of the Bias in the Neutron Multiplication Factor as a Function of Cooling Time, cp. Equation (3.3). (Initial Enrichment 4.0 wt.-%, Burnup 36.7 MWd/kg U, cp.

Figure 14).

1. Experimentalists should check their results on consistency as far as possible at all. The differences between measured and calculated data obtained for Eu-155 and Gd-155 from the GKN II samples for instance (cp. Figure 13) are not fully consistent. With these differences one gets negative number densities for Gd-155 at time of shut down.

2. One must admit, however, that chemical assay is hard to do, and so it is not surprising that one observes fluctuations in the differences between measured and calculated concentrations. For instance, even though the two GKN II samples are close together in initial enrichment and burnup the differences between measured and calculated mass ratios observed for the Sm isotopes are different in sign, cf. Figure 13. From the view of physics this means that there is no reason for revising the depletion code applied.

In addition, what matters with respect to burnup credit applications is the reactivity worth of the nuclides, since the actual point of interest is to estimate the uncertainty in the neutron

multiplication factor of a spent fuel management system of interest which is due to the uncertainties in the calculated isotopic concentrations. The result of a burnup credit criticality safety analysis is the determination of a loading curve for the spent fuel management system of interest. This curve specifies the loading criterion by indicating the minimum burnup necessary for the fuel assembly with a specific initial enrichment to be placed in the spent fuel management system designed for burnup credit. Reasonable loading curves provide for initial enrichments of 3.5 wt.-% or 3.8 wt.-% U-235 of course minimum burnup values which are significantly smaller than the burnup values of the GKN II samples. That means that the reactivity worth of the fission product concentrations referring to a loading curve is significantly smaller than the reactivity worth of the fission products in the GKN II samples. It follows, therefore, that the uncertainty in the neutron multiplication factor referring to the loading curve is smaller than that one which follows from the GKN II sample data.

3. FISSION PRODUCT REACTIVITY WORTH AND UNCERTAINTY IN THE