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Nuclear Engineering Division at Argonne
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Foreign Research Reactor Spent Nuclear Fuel

ANL/RERTR/TM-26

Table 2. MTR Fuel 93% Enrichment
MTR Fuel 93% Enrichment 100 g U-235
U-235 Burnup, % 0 5 10 20 30 40 50 60 70 80
U-235 Burned, g 0 5 10 20 30 40 50 60 70 80
U-234 0 0 0 0 0 0 0 0 0 0
U-235 100 95 90 80 70 60 50 40 30 20
U-236 0 1 2 3 5 6 8 9 11 12
U-238 8 8 8 8 7 7 7 7 7 7
U 108 103 99 91 82 74 65 56 48 39
Np-237 0 0.0 0.0 0.0 0.0 0.1 0.1 0.2 0.2 0.3
Np 0 0.0 0.0 0.0 0.0 0.1 0.1 0.2 0.2 0.3
Pu-238 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Pu-239 0 0.0 0.0 0.0 0.1 0.1 0.1 0.1 0.1 0.1
Pu-240 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Pu-241 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Pu-242 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Pu 0 0.0 0.0 0.0 0.1 0.1 0.1 0.1 0.2 0.2
Am-241 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Am 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
MTR Fuel 93% Enrichment 200 g U-235
U-235 Burnup, % 0 5 10 20 30 40 50 60 70 80
U-235 Burned, g 0 10 20 40 60 80 100 120 140 160
U-234 0 0 0 0 0 0 0 0 0 0
U-235 200 190 180 160 140 120 100 80 60 40
U-236 0 2 3 6 10 13 16 19 21 24
U-238 15 15 15 15 15 15 15 15 14 14
U 215 207 198 181 164 147 130 113 96 78
Np-237 0 0.0 0.0 0.0 0.1 0.2 0.3 0.4 0.6 0.8
Np 0 0.0 0.0 0.0 0.1 0.2 0.3 0.4 0.6 0.8
Pu-238 0 0.0 0.0 0.0 0.0 0.0 0.0 0.1 0.1 0.1
Pu-239 0 0.0 0.1 0.1 0.2 0.2 0.2 0.3 0.3 0.3
Pu-240 0 0.0 0.0 0.0 0.0 0.0 0.0 0.1 0.1 0.1
Pu-241 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Pu-242 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Pu 0 0.0 0.1 0.2 0.2 0.3 0.3 0.4 0.5 0.6
Am-241 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Am 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Table 2. MTR Fuel 93% Enrichment (conti.)
MTR Fuel 93% Enrichment 300 g U-235
U-235 Burnup, % 0 5 10 20 30 40 50 60 70 80
U-235 Burned, g 0 15 30 60 90 120 150 180 210 240
U-234 0 0 0 0 0 0 0 0 0 0
U-235 300 285 270 240 210 180 150 120 90 60
U-236 0 3 5 10 15 19 24 28 33 37
U-238 23 23 22 22 22 22 22 21 21 21
U 323 310 297 272 247 221 196 170 144 118
Np-237 0 0.0 0.0 0.1 0.2 0.4 0.6 0.8 1.1 1.5
Np 0 0.0 0.0 0.1 0.2 0.4 0.6 0.8 1.1 1.5
Pu-238 0 0.0 0.0 0.0 0.0 0.0 0.1 0.1 0.2 0.3
Pu-239 0 0.1 0.2 0.3 0.4 0.4 0.5 0.5 0.5 0.5
Pu-240 0 0.0 0.0 0.0 0.0 0.1 0.1 0.1 0.1 0.2
Pu-241 0 0.0 0.0 0.0 0.0 0.0 0.0 0.1 0.1 0.1
Pu-242 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Pu 0 0.1 0.2 0.3 0.4 0.5 0.7 0.8 0.9 1.1
Am-241 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Am 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
MTR Fuel 93% Enrichment 400 g U-235
U-235 Burnup, % 0 5 10 20 30 40 50 60 70 80
U-235 Burned, g 0 20 40 80 120 160 200 240 280 320
U-234 0 0 0 0 0 0 0 0 0 0
U-235 400 380 360 320 280 240 200 160 120 80
U-236 0 3 7 14 20 26 33 39 44 50
U-238 30 30 30 30 29 29 29 28 28 27
U 430 413 397 363 329 295 261 227 192 157
Np-237 0 0.0 0.0 0.2 0.4 0.6 0.9 1.3 1.7 2.2
Np 0 0.0 0.0 0.2 0.4 0.6 0.9 1.3 1.7 2.2
Pu-238 0 0.0 0.0 0.0 0.0 0.1 0.1 0.2 0.3 0.5
Pu-239 0 0.1 0.2 0.4 0.6 0.7 0.7 0.7 0.7 0.7
Pu-240 0 0.0 0.0 0.0 0.1 0.1 0.1 0.2 0.2 0.2
Pu-241 0 0.0 0.0 0.0 0.0 0.0 0.1 0.1 0.1 0.1
Pu-242 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.1
Pu 0 0.1 0.3 0.5 0.7 0.9 1.1 1.2 1.4 1.7
Am-241 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
Am 0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
2016 RERTR Meeting

The 2016 International RERTR Meeting (RERTR-2016) will take place in Belgium. Stay tuned for further details.

2015 RERTR Meeting

The 2015 International RERTR Meeting (RERTR-2015) took place in Seoul, Korea on Oct. 11-14, 2015.
For more information visit RERTR-2015.

ANL/RERTR/TM-26


ARGONNE NATIONAL LABORATORY, Nuclear Engineering Division, RERTR Department
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Last modified on July 29, 2008 11:33 +0200