Region II

Chapter 5. Application to spent fuel transportation and storage system for criticality control

5.1. Spent fuel storage pool

5.1.2. Region II

The region II racks have one neutron absorber between fuel assemblies. The adopted geometrical parameters are from Shin-kori units 3 and 4 as described in Table 15. The neutron absorber for Shin- kori units 3 and 4 is Boral with 0.0267 g B-10/cm for region II. In this study, the 75% credit for neutron absorber was applied, therefore, the concentration of B-10 used in the calculations is 0.0220 g B-10/cm for region II. The cell and sheathing are composed of stainless steel.

The fresh fuel condition of 5 wt% enriched UO2 fuel assembly was analyzed for optimization study in region II. The density of nuclear fuel is 10.313 g/cm (94.10TD). The water density is 1.0 g/cm which depicts maximum density. The neutron absorber is Boral and the B-10 areal density is 0.0165 g B-10/cm (75% neutron absorber credit). For conservative calculations, the soluble boron was not used in calculations. The boundary conditions are the periodic boundary for radial direction and vacuum boundary for axial direction. The keff for Boral was 1.20089 ± 0.00028.

The composition of Gd and B stainless steel is Gd 0.9 wt% + B 1.0 wt% with enriched B-10 90%. The optimization of inner width and thickness was performed for the inner width of 21.2 cm and 22.0 cm and thickness of 0.25 to 0.65 cm in Table 20. The reactivity for the inner width of 21.2 cm showed lower than that of 22.0 cm. The reactivity continuously decreases with increasing thickness of the cell wall.

However, the pitch could not be reduced with increasing thickness of the cell wall. Therefore, the pitch with optimized dimension and Gd and B stainless steel could be reduced from 22.5 cm to 22.05 cm as described in Figure 54.

The region II could store enrichment from 1.5 to 5.0 wt%. Therefore, the fresh fuel condition of each enrichment was calculated in Figure 55. Only enrichment of 1.5 wt% could be stored without burnup.

Therefore, the burnup credit is needed to be considered for storing over enrichment of 1.5 wt% for region II. The STARBUCS/SCALE code was used for depletion analysis. In STARBUCS calculation, the underneath burnup histories were considered.

 Average specific power: 38.365 MW/MTU

 No cooling time

 18 axial profiles from NUREG/CR-6801

 Nuclide sets selected from NUREG/CR-7109 (Actinide only, actinide with 16 fission products and all nuclides)

The nuclide sets are summarized in Table 21. For three nuclide sets, the minimum required burnups according to enrichment were calculated. The upper safety limit is set as 0.90 and 0.95. From all nuclide sets, the loading curves show consistency with Boral and Gd 0.9 wt% + B 1.0 wt% (enriched B-10) stainless steel with optimized dimensions. In other words, the Gd 0.9 wt% + B 1.0 wt% (enriched B- 10) stainless steel with optimized dimensions could be used in region II.

As described in Figure 53, there are thirteen modules of the rack in region II; four modules of 9x10, six modules of 10x10, two modules of 9x8 and one module of 9x9. The total number of fuel assemblies stored for Boral in region II is 1185. The increased capacity was calculated from Eq. (3)-(5). From the calculations, the capacity could be increased by about 4.05 %.

Table 15 Summary of dimensions of Shin-kori units 3 and 4

Component (cm) Region I Region II

Cell

Inner width 22.0 22.0

Thickness 0.25 0.25

Pitch 27.0 22.5

Neutron absorber

Width 18.0 18.0

Thickness 0.25 0.25

Sheathing Thickness 0.06 0.06

0.0 0.5 1.0 1.5 2.0 0.92

0.94 0.96 0.98 1.00 1.02

k

eff

Gd content (wt%)

B0.5 B0.8 B1.0 Boral

Figure 49 keff vs Gd content for B 0.5, 0.8 and 1.0 wt%

Table 16 Required neutron absorbing materials for the region I in Shin-kori units 3 and 4 B content (wt%) Gd content (wt%) Total (wt%)*

0.5 1.80 2.30 0.8 1.25 2.05 1.0 0.90 1.90

* Total amount of neutron absorbing materials (Gd+B)

0.5 0.6 0.7 0.8 0.9 1.0 0.8

1.0 1.2 1.4 1.6 1.8

Gd content (wt%)

B content (wt%)

Gd

y=2.70-1.80x

Figure 50 Required Gd contents as a function of B content for the region I in Shin-kori units 3 and 4 (x: B contents, y: Gd contents)

Table 17 keff vs cell wall thickness and cell inner width for the region I

Thickness (cm) Inner width (cm)

21.2 21.5 22.0

0.15 0.91381 0.92401 0.94028

0.25 0.89714 0.90971 0.92797

0.35 0.89105 0.90378 0.92534

0.45 0.88920 0.90241 0.92409

0.50 0.88952 0.90285 0.92483

25.0 25.5 26.0 26.5 27.0 0.88

0.90 0.92 0.94 0.96 0.98

Gd0.9B1.0 Boral

k

eff

Pitch (cm)

Figure 51 keff vs pitch for Gd 0.9 wt% + B 1.0 wt% stainless steel at cell inner width of 21.2 cm and the cell wall thickness of 0.45 cm

Table 18 keff vs cell wall thickness at cell inner width of 21.2 cm with enriched B-10 for the region I

Thickness (cm) Inner width (cm) 21.2

0.25 0.85320 0.35 0.84882 0.45 0.84762 0.50 0.84702 0.55 0.84748

24.4 24.5 24.6 24.7 24.8 24.9 25.0 25.1 0.92

0.93 0.94 0.95

Gd0.9B1.0 Boral

k

eff

Pitch (cm)

Figure 52 keff vs pitch for Gd 0.9 wt% + B 1.0 wt% with enriched B-10 90%

Table 19 keff calculated by KENO V.a. and MCS code for Gd 0.9 wt% + B 1.0 wt% with inner width of 21.2 cm, thickness of 0.50 cm and pitch of 24.9 cm

KENO V.a. MCS

0.92599±0.00033 0.92353±0.00031

Figure 53 Arrange of spent nuclear fuel for spent fuel storage pool in Shin-kori units 3 and 4 REGION I

REGION II

Table 20 keff vs cell wall thickness and cell inner width for the region II Thickness

(cm)

Inner width (cm)

21.2 22.0

0.25 1.18530 1.18780

0.50 1.15229 1.15388

0.55 1.14764 1.14946

0.60 1.14363 1.14463

0.65 1.14005 1.14115

21.8 22.0 22.2 22.4 22.6 1.14

1.15 1.16 1.17 1.18 1.19 1.20 1.21

k

eff

Pitch (cm)

Gd0.9B1.0_th0.25 Gd0.9B1.0_th0.55 Gd0.9B1.0_th0.60 Boral

Figure 54 keff vs pitch for Gd 0.9 wt% + B 1.0 wt% with enriched B-10 90% for the region II

1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 0.80

0.85 0.90 0.95 1.00 1.05 1.10 1.15 1.20 1.25

Gd0.9B1.0 Boral

k

eff

Enrichment (wt%)

Figure 55 keff vs enrichment in fresh fuel composition for Boral and Gd 0.9 wt% + B 1.0 wt%

Table 21 Nuclide sets used in burnup credit [35]

Set of nuclides for actinide-only (12)

U-234 U-235 U-236 U-238 Np-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-243

Set of nuclides for actinides and fission products (16)

Mo-95 Tc-99 Ru-101 Rh-103

Ag-109 Cs-133 Nd-143 Nd-145 Sm-147 Sm-149 Sm-150 Sm-151 Sm-152 Eu-151 Eu-153 Gd-155

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 0

10 20 30 40 50 60

Burnup (GWD/MTU)

Enrichment (wt%)

Gd0.9B1.0_USL0.90 Boral_USL0.90

Gd0.9B1.0_USL0.95 Boral_USL0.95

Figure 56 Loading curve for actinide only case

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 0

10 20 30 40 50 60

Burnup (GWD/MTU)

Enrichment (wt%)

Gd0.9B1.0_USL0.90 Boral_USL0.90

Gd0.9B1.0_USL0.95 Boral_USL0.95

Figure 57 Loading curve for actinide with 16 fission products case

1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 0

10 20 30 40 50 60

B u rnup ( G WD /MTU )

Enrichment (wt%)

Gd0.9B1.0_USL0.90 Boral_USL0.90

Gd0.9B1.0_USL0.95 Boral_USL0.95

Figure 58 Loading curve for all nuclides case