Argonne National Laboratory
Reduced Enrichment for Research and Test Reactors
Nuclear Science and Engineering Division at Argonne
U.S. Department of Energy

Home  :: Foreign Research Reactor Spent Nuclear Fuel :: Related Documents

RERTR Publications:
Foreign Research Reactor Spent Nuclear Fuel



In this example, a 280 g235U MTR-type fuel assembly has been irradiated at an average fuel assembly power () of 25 kW over an elapsed time () of 3584 days. The irradiation history of this fuel assembly is such that it can not be described simply, using a constant power () and a continuous irradiation time (). It is assumed, however, that

where the sum of () traces the fuel assembly irradiation history over all irradiation segments when the fuel assembly power was constant and the irradiation time was continuous. The elapsed time is the calendar time from the first through the last irradiation segment. Assuming 1.25 g235U burned per MWd, this fuel assembly has 112 g235U burned and 40% 235U burnup. The fission product decay time () or cooling time for this fuel assembly is assumed to be 3 years.

Nuclear Mass Inventory

If the fuel assembly enrichment is 93%, then 300 g235U, 40% 235U burnup data of Table 2 can be prorated to 280 g235U. For enrichments of 45 or 19.75%, similar prorated data from Table 3 or 4, respectively, should be used. Table D1 summarize the spent fuel mass inventory of 280 g235U fuel assemblies which have 40% 235U burnup.

Table D1. Mass Inventory of Spent HEU, MEU and LEU Fuel Assemblies
Isotope HEU-93% MEU-45% LEU-19.75%
U-234 0 0 0
U-235 168 168 168
U-236 18 18 19
U-238 21 337 1125
U 206 523 1311
Np-237 0.4 0.4 0.4
Np 0.4 0.4 0.4
Pu-238 0.0 0.0 0.0
Pu-239 0.4 3.2 7.0
Pu-240 0.1 0.6 1.1
Pu-241 0.0 0.2 0.4
Pu-242 0.0 0.0 0.0
Pu 0.5 3.9 8.6
Am-241 0.0 0.0 0.0
Am 0.0 0.0 0.0

These 280 g235U spent fuel inventory masses could also have been estimated using linear interpolation of the 200 and 300 g235U, 40% 235U burnup data tabulated in Tables 2, 3 and 4. Note, inventory masses for non-tabulated fuel assembly burnup should also use linear interpolation of tabulated data (e.g. 45% 235U burnup, interpolate between 40 and 50% tabulated data).

Photon Dose Rate

The photon dose rate of this fuel assembly is calculated from data presented in Table 8. The assembly power density is 0.089 MW/kg235U (25 kW / 280 g235U), the 235U burnup is 40%, and the decay time is 3 years. With these data, Table 8 estimates that the photon dose rate is 1.02 rem/h per g235U burned. With 112 g235U burned, the dose rate is 114 rem/h at 1 meter from the fuel assembly.

For fuel with 40% burnup and with 112 g235U burned, Fig. 1 estimates that this fuel assembly will be self-protecting (dose rate greater than 100 rem/h) for about 4 years.

The photon dose rate for non-tabulated assembly power densities, 235U burnup and/or decay times can be estimated using linear interpolation of the data in Table 8. Linear interpolation to determine the photon dose rate would be necessary, for example, for a fuel assembly with the following parameters: 3.5 year decay time, 50% 235U burnup and 0.134 MW/kg235U assembly power density. A simple table which interpolates each parameter separately is a useful aid. Table D2 is constructed to determine the photon dose rate for these non-tabulated fuel assembly parameters.

Table D2. Fuel Assembly Parameter Linear Interpolation
Decay Time, y Burnup, % 235U Assembly Power Density, MW/kg235U Photon Dose Rate, rem/h per g235U burned
3 50 0.179 1.31
3 50 0.089 1.07
3 50 0.134 1.19
4 50 0.179 1.10
4 50 0.089 0.931
4 50 0.134 1.0155
3.5 50 0.134 1.10

The bottom line, estimated photon dose rate is 1.10 rem/h per g235U burned.

Thermal Decay Heat

ORIGEN Calculation

The thermal decay heat calculated with the ORIGEN code for this example is about 4.2 Watts.

Integrated Emission Rate Equation

The thermal decay heat of this fuel assembly using the conservative heat load equation based upon Eq. -1


is about 10.6 W. This result is based upon an average fuel assembly power () of 25,000 Watts, a cooling or decay time () of 1095 days (3 y) and an elapsed time () of 3584 days.

El-Wakil Equation

The thermal decay heat with these same data and the heat load equation based upon Eq. -2


is about 5.1 W.

Untermyer and Weills Equation

Similarly, using the heat load equation based upon Eq. -3 with a decay time of 9.46107 seconds (1095 d) and an elapsed time of 3.10108 seconds (3584 d)


is about 3.8 W.

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.


ARGONNE NATIONAL LABORATORY, Nuclear Engineering Division, RERTR Department
9700 South Cass Ave., Argonne, IL 60439-4814
A U.S. Department of Energy laboratory managed by UChicago Argonne, LLC

Last modified on July 29, 2008 11:34 +0200