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R. B. Pond and J. E. Matos
Argonne National Laboratory
Argonne, IL


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This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies.

Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly.

Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg235U, and for fission product decay times of up to 20 years.

Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.


ANL/RERTR/TM-26 (original issue, May 1966) has been revised as follows:

Table 5 has been revised to increase the burnup range of TRIGA single-rod fuel from 35 to 60% 235U burnup.

Alternative thermal decay heat expressions have been included which are expected to give results close to actual heat loads of spent fuel. The previous thermal decay heat expression is expected to overestimate an actual heat load by about a factor of two. An analysis of the parameter constants used in the previous expression would suggest an uncertainty in calculated heat loads of the order of 10%. A decay heat comparison has been made for a typical fuel assembly using the ORIGEN code and the decay heat expressions.

Appendix C has been added which compares mass inventory estimates using the isotope generation and depletion code, ORIGEN and the cross section generation code, WIMS. Both codes solve material transmutation equations to determine material number densities. WIMS, however, solves the equations as a function of material burnup, while ORIGEN does not have a similar capability.

Appendix D has been added to illustrate an example calculation of the nuclear mass inventory, the photon dose rate, and the thermal decay heat for an assumed, spent MTR-type fuel assembly. All fuel assembly parameters necessary for the calculations are described.

Argonne National Laboratory, with facilities in the states of Illinois and Idaho, is owned by the United States government, and operated by The University of Chicago under the provisions of a contract with the Department of Energy.


This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

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