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# MOX Fuel Calculator - HELP

(last updated 27 Dec 2021)

Contents:

## Introduction

This calculator performs calculations of the material balance for Mixed Oxide (MOX) Fuel fabrication. This process combines plutonium recovered by reprocessing from spent fuel with a uranium component of various origins for use as fuel in light water reactors.

The material balance is presented in a flow chart. Upon entry of one value into any of the flow chart's input fields, all other fields are calculated accordingly. So, it is possible to calculate the balance per tonne of Plutonium Oxide PuO2, as well as per tonne of Heavy Metal (HM) in the Mixed Oxide (MOX) product, for example.
Note: any conversion losses between the various chemical forms of the materials are ignored.
Note: if U-236 is present in the uranium component, the actual concentration of U-235 and/or plutonium must be higher to compensate for the neutron toxicity of U-236.

The parameters used for the calculation can be set in the Process Parameters table. These parameters show reasonable initial values which can be modified as needed. There are no other hidden parameters used in the calculation.

Upon entry of any parameter, some MOX Product properties are calculated: the actual concentrations of the fissile plutonium (Puf) and the isotopes U-235, U-234 and U-236 in the Heavy Metal of the product.
The concentrations of U-234 and U-236 are also shown as microgram per gram U-235 (µg/gU-235). These figures are useful for comparison with the ASTM specifications (C 996-90) in LEU:

• 10,000 (or 11,000 by special agreement) µg/gU-235 for U-234, and
• 5,000 µg/gU-235 for U-236.
Exceedence of the ASTM specifications is indicated by yellow and/or red color flags.

## Process Parameters

Plutonium component
Puf equivalent of U-235 [g Puf per g U-235]
amount of fissile plutonium (Pu-239 and Pu-241) which produces the same energy in the reactor as 1 g of uranium-235.

Puf concentration in total Pu [wt-%]
concentration of fissile plutonium (Pu-239 and Pu-241) in the plutonium used for MOX fuel.

Uranium component
Sample
Selection of a sample parameter set initializes the uranium component parameters. After a selection is made, single parameters may be modified as required.
Assay [wt-% U-235] · [wt-% U-234] · [wt-% U-236]
Weight-percent of the isotopes uranium-235, uranium-234 and uranium-236 in the uranium component used for MOX fuel production.
U-235 is contained in natural uranium at 0.711 wt-%, and in depleted uranium typically at 0.2 - 0.35 wt-%. Slightly enriched and reprocessed uranium contain U-235 at various assays.
U-234 is a minor isotope contained in natural uranium. Caution: Upon entry of a value for the U-235 assay, or the Uranium component Origin or Tails Assay (see below), the calculator automatically determines a value for the U-234 assay. In case the estimated value is inappropriate, it can be overwritten.
Note: For depleted uranium as uranium component it is assumed that it is a by-product from enrichment of natural uranium to 4% U-235, if no other product assay is entered below.
U-236 is not found in natural uranium, but any processed uranium may contain traces of U-236, depending on its manufacturing history.
Origin Assay · Tails Assay [wt-% U-235]   (Only required if Uranium component is obtained by enrichment of Uranium component Origin)
Weight-percent of uranium-235 in the uranium originally used as feed resp. generated as tails in the enrichment process that supplied the Uranium component.
Note: In case the Uranium component Assay is lower than the Origin Assay - that is DU is used for Uranium component, the product assay for the enrichment process (in which the DU was generated) can be entered instead of the tails assay. Otherwise a product assay of 4% is assumed.

MOX Product
Assay [wt-% U-235 equiv.]
Concentration of fissile plutonium and uranium-235 in the MOX product, expressed as weight-percent equivalent of U-235.
(The actual assays may be higher to take the neutron-absorbing effect of U-236 into account.)

General
Factor for U-236 effect [excess wt-% U-235 per wt-% U-236]
Excess concentration of U-235 required to offset for the neutron-absorbing effect of U-236 present in the MOX product.
Typical values are in the 0.2 - 0.3 range, depending on reactor and fuel type.

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