Fig. 1.
Radioactive decay over time of spent CANDU fuel upon discharge from the reactor. It uses natural uranium in comparison and shows that it will take up to 10 million years before the spent fuel goes back to the same overall radioactivity levels as the original ore equivalent. The same chart can be applied qualitatively to any kind of uranium-based SF.
Fig. 2.
Decay of VHLW from reprocessing of 1 ton of spent fuel [4].
Fig. 3.
Radiological dose rate in Sv/h from two French HLW drill cores. C2 is VHLW; CU1 is SF (p. 25 in Ref. [2] and p. 533 in Ref. [7]).
Fig. 4.
Accrued, external dose from drill cores of several kinds of present and future French VHLW after a 10-min exposure, 40 cm from the drill core, and as a function of time [9].
Table 1.
Gamma dose rate (mSv/h) at three distances from a POLLUX-8 cask loaded with 8 UO2 fuel assemblies, neglecting the shielding of the cask
| Decay time (million years) | Next to surface (4 cm) | 40 cm (extrapolated) | 10 m |
|---|---|---|---|
| 1 | 6.56 | 0.697 | 0.0279 |
| 10 | 2.77 | 0.290 | 0.0116 |
| 100 | 2.68 | 0.282 | 0.0113 |
| 1000 | 2.31 | 0.243 | 0.00974 |
Table 2a.
Main contributors to the gamma dose rate from SF as a function of time 4 in million years for the configuration of Table 1. The (4n + 2) elements are denoted in bold and the (4n + 1) elements are denoted in italics
| 0.1 My | 1 My | 10 My | 100 My | 1000 My |
|---|---|---|---|---|
| Bi-214 | Bi-214 | Bi-214 | Bi-214 | Bi-214 |
| Sb-126m | Pa-233 | Pb-214 | Pb-214 | Pb-214 |
| Pb-214 | Th-229 | Pa-234m | Pa-234m | Pa-234m |
| Pa-233 | Bi-213 | Pa-233 | Bi-210 | Bi-210 |
| Sb-126 | Pb-214 | Th-229 | U-235 | Th-234 |
| Th-229 | Tl-209 | Bi-213 | Ra-223 | |
| Np-237 | Np-237 | Bi-210 | ||
| Bi-213 | Fr-221 | U-235 | ||
| Bi-210 | Ac-225 | Ra-223 | ||
| Tl-209 | Ra-225 | Tl-209 |
Table 2b.
Highest gamma energies in the (4n + 2) and (4n + 1) chains. The weighted averages values are based on the data obtained from Japan Atomic Energy Agency [15,16,17,18]
| Radionuclide | Weighted average gamma energy per emission (keV) |
|---|---|
| Bi-214 | 944.00 |
| Pb-214 | 318.38 |
| Pa-233 | 311.86 |
| Th-229 | 47.96 |
Table 3.
Relative contributions from the three actinide chains to the gamma dose from a block of unshielded eight UO2 fuel assemblies (same assemblies as mentioned in Table 1)
| Decay time (million years) | U-238 chain (%) | Np-237 chain (%) | Excess U-234 chain (%) | Total, surface dose rate (mSv/h) |
|---|---|---|---|---|
| 1 | 41.12 | 32.10 | 26.78 | 6.56 |
| 2 | 62.41 | 35.16 | 2.41 | 4.32 |
| 4 | 77.28 | 22.71 | – | 3.49 |
| 6 | 92.31 | 7.69 | – | 2.92 |
| 8 | 95.73 | 4.27 | – | 2.81 |
| 10 | 96.29 | 3.72 | – | 2.77 |
| 14 | 98.86 | 1.14 | – | 2.72 |
| 20 | 99.83 | 0.17 | – | 2.69 |
| 100 | 100 | – | – | 2.68 |
| 1000 | 100 | – | – | 2.31 |
Table 4.
Exposure time in hours for getting the annual ICRP dose constraint of 0.3 mSv as a function of time at three distances from POLLUX-8 cask assemblies, neglecting the shielding of the cask
| Decay time (million years) | Exposure (h) at contact | Exposure (h) at 40 cm | Exposure (h) at 10 m |
|---|---|---|---|
| 1 | 0.046 | 0.430 | 10.8 |
| 10 | 0.108 | 1.034 | 25.9 |
| 100 | 0.112 | 1.064 | 26.6 |
| 1000 | 0.130 | 1.234 | 30.8 |
Table 5.
Elemental activity ratios (at 10 000 years) and chain activity ratios (at one million years) for French VHLW
| Time (years) | Np-237/(U-234 + U-238) | Np-237/U-238 | U-234/U-238 |
|---|---|---|---|
| 10 000 | 576.73 | 19 220.62 | 32.34 |
| 1 000 000 | 4672.93 | 13 896.10 | 1.97 |
Fig. 5.
The dose rate 2 cm away from the full-size C2 VHLW cylinder as a function of time (source: ANDRA).
Table A1.1.
Density and HVL for various materials depending on the energy of the gamma rays
| Material | Density (g/cm3) | HVL (cm) for 0.3 MeV | HVL (cm) for 0.6 MeV | HVL (cm) for 1.0 MeV | HVL (cm) for 1.5 MeV |
|---|---|---|---|---|---|
| Concrete | 2.3 | 2.08 | 4.02 | 6.69 | 7.70 |
| R7T7 glass | 2.6 | 2.22 | 4.10 | 5.92 | 6.50 |
| UO2 | 10.97 | 0.26 | 0.49 | 0.79 | 1.33 |
Table A1.2.
Transmission factors for energies 0.944 MeV and 0.312 MeV in R7T7 glass waste forms of radii 21 cm and 5 cm
| Energy (MeV) | Cylinder radius (cm) | HVL (cm) | Transmission factor (%) |
|---|---|---|---|
| 0.944 | 21 | 3.757 | 8.53 |
| 0.312 | 21 | 1.576 | 0.28 |
| 0.944 | 5 | 3.757 | 56.80 |
| 0.312 | 5 | 1.576 | 27.90 |
Table A2.1.
Dose rate computation coefficients for the German spent fuel of Table 1
| Ai (mSv/h) | λi (1/106 years) | |
|---|---|---|
| U-238 | 2.697 | 1.55 × 10−4 |
| Np-237 | 2.910 | 0.324 |
| U-234 | 29.670 | 2.820 |